High-heat-flux testing of helium-cooled heat exchangers for fusion applications

International Nuclear Information System (INIS)

Youchison, D.L.; Izenson, M.G.; Baxi, C.B.; Rosenfeld, J.H.

1996-01-01

High-heat-flux experiments on three types of helium-cooled divertor mock-ups were performed on the 30-kW electron beam test system and its associated helium flow loop at Sandia National Laboratories. A dispersion-strengthened copper alloy (DSCu) was used in the manufacture of all the mock-ups. The first heat exchanger provides for enhanced heat transfer at relatively low flow rates and much reduced pumping requirements. The Creare sample was tested to a maximum absorbed heat flux of 5.8 MW/m 2 . The second used low pressure drops and high mass flow rates to achieve good heat removal. The GA specimen was tested to a maximum absorbed heat flux of 9 MW/m 2 while maintaining a surface temperature below 400 degree C. A second experiment resulted in a maximum absorbed heat flux of 34 MW/m 2 and surface temperatures near 533 degree C. The third specimen was a DSCu, axial flow, helium-cooled divertor mock-up filled with a porous metal wick which effectively increases the available heat transfer area. Low mass flow and high pressure drop operation at 4.0 MPa were characteristic of this divertor module. It survived a maximum absorbed heat flux of 16 MW/m 2 and reached a surface temperature of 740 degree C. Thermacore also manufactured a follow-on, dual channel porous metal-type heat exchanger, which survived a maximum absorbed heat flux of 14 MW/m 2 and reached a maximum surface temperature of 690 degree C. 11refs., 20 figs., 3 tabs

Elevated temperature and high pressure large helium gas loop

International Nuclear Information System (INIS)

Sakasai, Minoru; Midoriyama, Shigeru; Miyata, Toyohiko; Nakase, Tsuyoshi; Izaki, Makoto

1979-01-01

The development of high temperature gas-cooled reactors especially aiming at the multi-purpose utilization of nuclear heat energy is carried out actively in Japan and West Germany. In Japan, the experimental HTGR of 50 MWt and 1000 deg C outlet temperature is being developed by Japan Atomic Energy Research Institute and others since 1969, and the development of direct iron-making technology utilizing high temperature reducing gas was started in 1973 as the large project of Ministry of Internalional Trade and Industry. Kawasaki Heavy Industries, Ltd., Has taken part in these development projects, and has developed many softwares for nuclear heat design, system design and safety design of nuclear reactor system and heat utilization system. In hardwares also, efforts have been exerted to develop the technologies of design and manufacture of high temperature machinery and equipments. The high temperature, high pressure, large helium gas loop is under construction in the technical research institute of the company, and it is expected to be completed in December, 1979. The tests planned are that of proving the dynamic performances of the loop and its machinery and equipments and the verification of analysis codes. The loop is composed of the main circulation system, the objects of testing, the helium gas purifying system, the helium supplying and evacuating system, instruments and others. (Kako, I.)

Helium circulator design concepts for the modular high temperature gas-cooled reactor (MHTGR) plant

International Nuclear Information System (INIS)

McDonald, C.F.; Nichols, M.K.; Kaufman, J.S.

1988-01-01

Two helium circulators are featured in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) power plant - (1) the main circulator, which facilitates the transfer of reactor thermal energy to the steam generator, and (2) a small shutdown cooling circulator that enables rapid cooling of the reactor system to be realized. The 3170 kW(e) main circulator has an axial flow compressor, the impeller being very similar to the unit in the Fort St. Vrain (FSV) plant. The 164 kW(e) shutdown cooling circulator, the design of which is controlled by depressurized conditions, has a radial flow compressor. Both machines are vertically oriented, have submerged electric motor drives, and embody rotors that are supported on active magnetic bearings. As outlined in this paper, both machines have been conservatively designed based on established practice. The circulators have features and characteristics that have evolved from actual plant operating experience. With a major goal of high reliability, emphasis has been placed on design simplicity, and both machines are readily accessible for inspection, repair, and replacement, if necessary. In this paper, conceptual design aspects of both machines are discussed, together with the significant technology bases. As appropriate for a plant that will see service well into the 21st century, new and emerging technologies have been factored into the design. Examples of this are the inclusion of active magnetic bearings, and an automated circulator condition monitoring system. (author). 18 refs, 20 figs, 13 tabs

Radiolytic reactions in the coolant of helium cooled reactors

International Nuclear Information System (INIS)

Tingey, G.L.; Morgan, W.C.

1975-01-01

The success of helium cooled reactors is dependent upon the ability to prevent significant reaction between the coolant and the other components in the reactor primary circuit. Since the thermal reaction of graphite with oxidizing gases is rapid at temperatures of interest, the thermal reactions are limited primarily by the concentration of impurity gases in the helium coolant. On the other hand, the rates of radiolytic reactions in helium are shown to be independent of reactive gas concentration until that concentration reaches a very low level. Calculated steady-state concentrations of reactive species in the reactor coolant and core burnoff rates are presented for current U. S. designed, helium cooled reactors. Since precise base data are not currently available for radiolytic rates of some reactions and thermal reaction rate data are often variable, the accuracy of the predicted gas composition is being compared with the actual gas compositions measured during startup tests of the Fort Saint Vrain high temperature gas-cooled reactor. The current status of these confirmatory tests is discussed. 12 references

Corrosion of high temperature alloys in the primary circuit helium of high temperature gas cooled reactors. Pt. 2

International Nuclear Information System (INIS)

Quadakkers, W.J.

1985-01-01

The reactive impurities H 2 O, CO, H 2 and CH 4 which are present in the primary coolant helium of high temperature gas-cooled reactors can cause scale formation, internal oxidation and carburization or decarburization of the high temperature structural alloys. In Part 1 of this contribution a theoretical model was presented, which allows the explanation and prediction of the observed corrosion effects. The model is based on a classical stability diagram for chromium, modified to account for deviations from equilibrium conditions caused by kinetic factors. In this paper it is shown how a stability diagram for a commercial alloy can be constructed and how this can be used to correlate the corrosion results with the main experimental parameters, temperature, gas and alloy composition. Using the theoretical model and the presented experimental results, conditions are derived under which a protective chromia based surface scale will be formed which prevents a rapid transfer of carbon between alloy and gas atmosphere. It is shown that this protective surface oxide can only be formed if the carbon monoxide pressure in the gas exceeds a critical value. Psub(CO), which depends on temperature and alloy composition. Additions of methane only have a limited effect provided that the methane/water ratio is not near to, or greater than, a critical value of around 100/1. The influence of minor alloying additions of strong oxide forming elements, commonly present in high temperature alloys, on the protective properties of the chromia surface scales and the kinetics of carbon transfer is illustrated. (orig.) [de

Summary report on technical experiences from high-temperature helium turbomachinery testing in Germany

International Nuclear Information System (INIS)

Weisbrodt, I.A.

1996-01-01

In Germany a comprehensive research and development program was initiated in 1968 for a Brayton (closed) cycle power conversion system. The program was for ultimate use with a high temperature, helium cooled nuclear reactor heat source (the HHT project) for electricity generation using helium as the working fluid. The program continued until 1982 in international cooperation with the United States and Switzerland. This document describes the designs and reports the results of testing activities that addressed the development of turbines, compressors, hot gas ducts, materials, heat exchangers and other equipment items for use with a helium working fluid at high temperatures. 67 refs, 34 figs, tabs

High-temperature helium-loop facility

International Nuclear Information System (INIS)

Tokarz, R.D.

1981-09-01

The high-temperature helium loop is a facility for materials testing in ultrapure helium gas at high temperatures. The closed loop system is capable of recirculating high-purity helium or helium with controlled impurities. The gas loop maximum operating conditions are as follows: 300 psi pressure, 500 lb/h flow rate, and 2100 0 F temperature. The two test sections can accept samples up to 3.5 in. diameter and 5 ft long. The gas loop is fully instrumented to continuously monitor all parameters of loop operation as well as helium impurities. The loop is fully automated to operate continuously and requires only a daily servicing by a qualified operator to replenish recorder charts and helium makeup gas. Because of its versatility and high degree of parameter control, the helium loop is applicable to many types of materials research. This report describes the test apparatus, operating parameters, peripheral systems, and instrumentation system. The experimental capabilities and test conand presents the results that have been obtained. The study has been conducted using a four-phase approach. The first phase develops the solution to the steady-state radon-diffusion equation in one-dimensieered barriers; disposal charge analysis; analysis of spent fuel policy implementation; spent f water. Field measurements and observations are reported for each site. Analytical data and field measurements are presented in tables and maps. Uranium concentrations in the sediments which were above detection limits ranged from 0.10 t 51.2 ppM. The mean of the logarithms of the uranium concentrations was 0.53. A group of high uranium concentrations occurs near the junctions of quadrangles AB, AC, BB, a 200 mK. In case 2), x-ray studies of isotopic phase separation in 3 He-- 4 He bcc solids were carried out by B. A. Fraass

The R&D of HTGR high temperature helium sampling loop: From HTR-10 to HTR-PM

Energy Technology Data Exchange (ETDEWEB)

Fang, Chao, E-mail: fangchao@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Collaborative Innovation Center of Advanced Nuclear Energy Technology, Tsinghua University, Beijing 100084 (China); The Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing 100084 (China); Bao, Xuyin; Yang, Chen; Yang, Yanran; Cao, Jianzhu [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Collaborative Innovation Center of Advanced Nuclear Energy Technology, Tsinghua University, Beijing 100084 (China); The Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing 100084 (China)

2016-09-15

A High Temperature Helium Sampling Loop (HTHSL) for studying the transportation (deposition) behavior and total amount of solid fission products in high-temperature helium coming from the steam generator (SG) in the 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) and High Temperature Reactor-Pebble bed Modules (HTR-PM) are researched and designed, respectively. Through the optimal design and simulation based on thermohydraulics analysis, the three-sleeve structure of deposition sampling device (DSD) could realize full-length temperature control evenly so that it could be used to study fission products in the primary circuit of HTR-10. On the other hand, an improved DSD is also designed for HTR-PM based on corresponding simulations, which could be used to sample the important nuclei in the high temperature helium from SG. These schemes offer two different methods to obtain the original source term in the high temperature helium, which will provide deeper understanding for the analysis of source terms of HTGR.

Helium cooling of fusion reactors

International Nuclear Information System (INIS)

Wong, C.P.C.; Baxi, C.; Bourque, R.; Dahms, C.; Inamati, S.; Ryder, R.; Sager, G.; Schleicher, R.

1994-01-01

On the basis of worldwide design experience and in coordination with the evolution of the International Thermonuclear Experimental Reactor (ITER) program, the application of helium as a coolant for fusion appears to be at the verge of a transition from conceptual design to engineering development. This paper presents a review of the use of helium as the coolant for fusion reactor blanket and divertor designs. The concept of a high-pressure helium cooling radial plate design was studied for both ITER and PULSAR. These designs can resolve many engineering issues, and can help with reaching the goals of low activation and high performance designs. The combination of helium cooling, advanced low-activation materials, and gas turbine technology may permit high thermal efficiency and reduced costs, resulting in the environmental advantages and competitive economics required to make fusion a 21st century power source. ((orig.))

Evaluation of helium cooling for fusion divertors

International Nuclear Information System (INIS)

Baxi, C.B.

1993-09-01

The divertors of future fusion reactors will have a power throughput of several hundred MW. The peak heat flux on the diverter surface is estimated to be 5 to 15 MW/m 2 at an average heat flux of 2 MW/m 2 . The divertors have a requirement of both minimum temperature (100 degrees C) and maximum temperature. The minimum temperature is dictated by the requirement to reduce the absorption of plasma, and the maximum temperature is determined by the thermo-mechanical properties of the plasma facing materials. Coolants that have been considered for fusion reactors are water, liquid metals and helium. Helium cooling has been shown to be very attractive from safety and other considerations. Helium is chemically and neutronically inert and is suitable for power conversion. The challenges associated with helium cooling are: (1) Manifold sizes; (2) Pumping power; and (3) Leak prevention. In this paper the first two of the above design issues are addressed. A variety of heat transfer enhancement techniques are considered to demonstrate that the manifold sizes and the pumping power can be reduced to acceptable levels. A helium-cooled diverter module was designed and fabricated by GA for steady-state heat flux of 10 MW/m 2 . This module was recently tested at Sandia National Laboratories. At an inlet pressure of 4 MPa, the module was tested at a steady-state heat flux of 10 MW/m 2 . The pumping power required was less than 1% of the power removed. These results verified the design prediction

An efficient continuous flow helium cooling unit for Moessbauer experiments

International Nuclear Information System (INIS)

Herbert, I.R.; Campbell, S.J.

1976-01-01

A Moessbauer continuous flow cooling unit for use with liquid helium over the temperature range 4.2 to 300K is described. The cooling unit can be used for either absorber or source studies in the horizontal plane and it is positioned directly on top of a helium storage vessel. The helium transfer line forms an integral part of the cooling unit and feeds directly into the storage vessel so that helium losses are kept to the minimum. The helium consumption is 0.12 l h -1 at 4.2 K decreasing to 0.055 l h -1 at 40 K. The unit is top loading and the exchange gas cooled samples can be changed easily and quickly. (author)

HTGR [High Temperature Gas-Cooled Reactor] ingress analysis using MINET

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Yang, J.W.; Kroeger, P.G.; Mallen, A.N.; Aronson, A.L.

1989-04-01

Modeling of water/steam ingress into the primary (helium) cooling circuit of a High Temperature Gas-Cooled Reactor (HTGR) is described. This modeling was implemented in the MINET Code, which is a program for analyzing transients in intricate fluid flow and heat transfer networks. Results from the simulation of a water ingress event postulated for the Modular HTGR are discussed. 27 refs., 6 figs., 6 tabs

Concept Design for a High Temperature Helium Brayton Cycle with Interstage Heating and Cooling

Energy Technology Data Exchange (ETDEWEB)

Wright, Steven A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vernon, Milton E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pickard, Paul S. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2013-12-01

The primary metric for the viability of these next generation nuclear power plants will be the cost of generated electricity. One important component in achieving these objectives is the development of power conversion technologies that maximize the electrical power output of these advanced reactors for a given thermal power. More efficient power conversion systems can directly reduce the cost of nuclear generated electricity and therefore advanced power conversion cycle research is an important area of investigation for the Generation IV Program. Brayton cycles using inert or other gas working fluids, have the potential to take advantage of the higher outlet temperature range of Generation IV systems and allow substantial increases in nuclear power conversion efficiency, and potentially reductions in power conversion system capital costs compared to the steam Rankine cycle used in current light water reactors. For the Very High Temperature Reactor (VHTR), Helium Brayton cycles which can operate in the 900 to 950 C range have been the focus of power conversion research. Previous Generation IV studies examined several options for He Brayton cycles that could increase efficiency with acceptable capital cost implications. At these high outlet temperatures, Interstage Heating and Cooling (IHC) was shown to provide significant efficiency improvement (a few to 12%) but required increased system complexity and therefore had potential for increased costs. These scoping studies identified the potential for increased efficiency, but a more detailed analysis of the turbomachinery and heat exchanger sizes and costs was needed to determine whether this approach could be cost effective. The purpose of this study is to examine the turbomachinery and heat exchanger implications of interstage heating and cooling configurations. In general, this analysis illustrates that these engineering considerations introduce new constraints to the design of IHC systems that may require

Study of Temperature Wave Propagation in Superfluid Helium Focusing on Radio-Frequency Cavity Cooling

CERN Document Server

Koettig, T; Avellino, S; Junginger, T; Bremer, J

2015-01-01

Oscillating Superleak Transducers (OSTs) can be used to localize quenches of superconducting radio-frequency cavities. Local hot spots at the cavity surface initiate temperature waves in the surrounding superfluid helium that acts as cooling fluid at typical temperatures in the range of 1.6 K to 2 K. The temperature wave is characterised by the properties of superfluid helium such as the second sound velocity. For high heat load densities second sound velocities greater than the standard literature values are observed. This fast propagation has been verified in dedicated small scale experiments. Resistors were used to simulate the quench spots under controlled conditions. The three dimensional propagation of second sound is linked to OST signals. The aim of this study is to improve the understanding of the OST signal especially the incident angle dependency. The characterised OSTs are used as a tool for quench localisation on a real size cavity. Their sensitivity as well as the time resolution was proven to b...

LOFA analyses for the water and helium cooled SEAFP reactors

International Nuclear Information System (INIS)

Sponton, L.; Sjoeberg, A.; Nordlinder, S.

2001-01-01

This study was performed in the frame of the European long-term fusion safety programme 1999 (SEAFP99). Loss of flow accidents (LOFA) have been studied for two cases, first for a helium cooled reactor with advanced dual-coolant (DUAL) blanket at 100% nominal power. The second case applies to a water-cooled reactor at 20% nominal power. Both transients were simulated with the code MELCOR 1.8.4. The results for the helium cooled reactor show that with a natural circulation flow of helium after the pump stops, the first wall temperature will stay below the temperature for excepted failure of the construction material. For the water cooled reactor, the results show that the pressurizer set point for its liquid volumetric inventory is reached before the plasma facing components attain a critical temperature. The pressurizer set point will induce a plasma shutdown

Thermal Performance of a Dual-Channel, Helium-Cooled, Tungsten Heat Exchanger

International Nuclear Information System (INIS)

Youchison, Dennis L.; North, Mart T.

2000-01-01

Helium-cooled, refractory heat exchangers are now under consideration for first wall and divertor applications. These refractory devices take advantage of high temperature operation with large delta-Ts to effectively handle high heat fluxes. The high temperature helium can then be used in a gas turbine for high-efficiency power conversion. Over the last five years, heat removal with helium was shown to increase dramatically by using porous metal to provide a very large effective surface area for heat transfer in a small volume. Last year, the thermal performance of a bare-copper, dual-channel, helium-cooled, porous metal divertor mock-up was evaluated on the 30 kW Electron Beam Test System at Sandia National Laboratories. The module survived a maximum absorbed heat flux of 34.6 MW/m 2 and reached a maximum surface temperature of 593 C for uniform power loading of 3 kW absorbed on a 2-cm 2 area. An impressive 10 kW of power was absorbed on an area of 24 cm 2 . Recently, a similar dual-module, helium-cooled heat exchanger made almost entirely of tungsten was designed and fabricated by Thermacore, Inc. and tested at Sandia. A complete flow test of each channel was performed to determine the actual pressure drop characteristics. Each channel was equipped with delta-P transducers and platinum RTDs for independent calorimetry. One mass flow meter monitored the total flow to the heat exchanger, while a second monitored flow in only one of the channels. The thermal response of each tungsten module was obtained for heat fluxes in excess of 5 MW/m 2 using 50 C helium at 4 MPa. Fatigue cycles were also performed to assess the fracture toughness of the tungsten modules. A description of the module design and new results on flow instabilities are also presented

Mechanical characterization of metallic materials for high-temperature gas-cooled reactors in air and in helium environments

International Nuclear Information System (INIS)

Sainfort, G.; Cappelaere, M.; Gregoire, J.; Sannier, J.

1984-01-01

In the French R and D program for high-temperature gas-cooled reactors (HTGRs), three metallic alloys were studied: steel Chromesco-3 with 2.25% chromium, alloy 800H, and Hastelloy-X. The Chromesco-3 and alloy 800H creep behavior is the same in air and in HTGR atmosphere (helium). The tensile tests of Hastelloy-X specimens reveal that aging has embrittlement and hardening effects up to 700 0 C, but the creep tests at 800 0 C show opposite effects. This particular behavior could be due to induced precipitation by aging and the depletion of hardening elements from the matrix. Tests show a low influence of cobalt content on mechanical properties of Hastelloy-X

Options for a high heat flux enabled helium cooled first wall for DEMO

Energy Technology Data Exchange (ETDEWEB)

Arbeiter, Frederik, E-mail: f.arbe@kit.edu; Chen, Yuming; Ghidersa, Bradut-Eugen; Klein, Christine; Neuberger, Heiko; Ruck, Sebastian; Schlindwein, Georg; Schwab, Florian; Weth, Axel von der

2017-06-15

Highlights: • Design challenges for helium cooled first wall reviewed and otimization approaches explored. • Application of enhanced heat transfer surfaces to the First Wall cooling channels. • Demonstrated a design point for 1 MW/m{sup 2} with temperatures <550 °C and acceptable stresses. • Feasibility of several manufacturing processes for ribbed surfaces is shown. - Abstract: Helium is considered as coolant in the plasma facing first wall of several blanket concepts for DEMO fusion reactors, due to the favorable properties of flexible temperature range, chemical inertness, no activation, comparatively low effort to remove tritium from the gas and no chemical corrosion. Existing blanket designs have shown the ability to use helium cooled first walls with heat flux densities of 0.5 MW/m{sup 2}. Average steady state heat loads coming from the plasma for current EU DEMO concepts are expected in the range of 0.3 MW/m{sup 2}. The definition of peak values is still ongoing and depends on the chosen first wall shape, magnetic configuration and assumptions on the fraction of radiated power and power fall off lengths in the scrape off layer of the plasma. Peak steady state values could reach and excess 1 MW/m{sup 2}. Higher short-term transient loads are expected. Design optimization approaches including heat transfer enhancement, local heat transfer tuning and shape optimization of the channel cross section are discussed. Design points to enable a helium cooled first wall capable to sustain heat flux densities of 1 MW/m{sup 2} at an average shell temperature lower than 500 °C are developed based on experimentally validated heat transfer coefficients of structured channel surfaces. The required pumping power is in the range of 3–5% of the collected thermal power. The FEM stress analyses show code-acceptable stress intensities. Several manufacturing methods enabling the application of the suggested heat transfer enhanced first wall channels are explored. An

Cooling with Superfluid Helium

Energy Technology Data Exchange (ETDEWEB)

Lebrun, P; Tavian, L [European Organization for Nuclear Research, Geneva (Switzerland)

2014-07-01

The technical properties of helium II (‘superfluid’ helium) are presented in view of its applications to the cooling of superconducting devices, particularly in particle accelerators. Cooling schemes are discussed in terms of heat transfer performance and limitations. Large-capacity refrigeration techniques below 2 K are reviewed, with regard to thermodynamic cycles as well as process machinery. Examples drawn from existing or planned projects illustrate the presentation. Keywords: superfluid helium, cryogenics.

Thermodynamic properties of helium in the range from 20 to 15000C and 1 to 100 bar. Reactor core design of high-temperature gas-cooled reactors. Pt. 1

International Nuclear Information System (INIS)

Kipke, H.E.; Stoehr, A.; Banerjea, A.; Hammeke, K.; Huepping, N.

1978-12-01

The following report presents in tabular form the safety standard of the nuclear safety standard commission (KTA) on reactor core design of high-temperature gas-cooled reactors. Part 1: Calculation of thermodynamic properties of helium The basis of the present work is the data and formulae given by H. Petersen for the calculation of density, specific heat, thermal conductivity and dynamic viscosity of helium together with the formula for their standard deviations in the range of temperature and pressure stated above. The relations for specific enthalpy and specific entropy have been derived from density and specific heat, whereby specific heat is assumed constant over the given range of temperature and pressure. The latter section of this report contains tables of thermodynamic properties of helium calculated from the equations stated earlier in this paper. (orig.) [de

Parametric Investigation of Brayton Cycle for High Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Chang Oh

2004-01-01

The Idaho National Engineering and Environmental Laboratory (INEEL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. In this project, we are investigating helium Brayton cycles for the secondary side of an indirect energy conversion system. Ultimately we will investigate the improvement of the Brayton cycle using other fluids, such as supercritical carbon dioxide. Prior to the cycle improvement study, we established a number of baseline cases for the helium indirect Brayton cycle. These cases look at both single-shaft and multiple-shaft turbomachinery. The baseline cases are based on a 250 MW thermal pebble bed HTGR. The results from this study are applicable to other reactor concepts such as a very high temperature gas-cooled reactor (VHTR), fast gas-cooled reactor (FGR), supercritical water reactor (SWR), and others. In this study, we are using the HYSYS computer code for optimization of the helium Brayton cycle. Besides the HYSYS process optimization, we performed parametric study to see the effect of important parameters on the cycle efficiency. For these parametric calculations, we use a cycle efficiency model that was developed based on the Visual Basic computer language. As a part of this study we are currently investigated single-shaft vs. multiple shaft arrangement for cycle efficiency and comparison, which will be published in the next paper. The ultimate goal of this study is to use supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency to values great than that of the helium Brayton cycle. This paper includes preliminary calculations of the steady state overall Brayton cycle efficiency based on the pebble bed reactor reference design (helium used as the working fluid) and compares those results with an initial calculation of a CO2 Brayton cycle

Conceptual design of two helium cooled fusion blankets (ceramic and liquid breeder) for INTOR

International Nuclear Information System (INIS)

Dalle Donne, M.; Dorner, S.; Taczanowski, S.

1983-08-01

Neutronic and heat transfer calculations have been performed for two helium cooled blankets for the INTOR design. The neutronic calculations show that the local tritium breeding ratios, both for the ceramic blanket (Li 2 SiO 3 ) and for the liquid blanket (Li 17 Pb 83 ) solutions, are 1.34 for natural tritium and about 1.45 using 30% Li 6 enrichment. The heat transfer calculations show that it is possible to cool the divertor section of the torus (heat flux = 1.7 MW/m 2 ) with helium with an inlet pressure of 52 bar and an inlet temperature of 40 0 C. The temperature of the back face of the divertor can be kept at 130 0 C. With helium with the same inlet conditions it is possible to cool the first wall as well (heat flux = 0.136 MW/m 2 ) and keep the back-face of this wall at a temperature of 120 0 C. For the ceramic blanket we use helium with 52 bar inlet pressure and 400 0 C inlet temperature to ensure sufficiently high temperatures in the breeder material. The maximum temperature in the pressure tubes containing the blanket is 450 0 C, while the maximum breeder particle temperature is 476 0 C. (orig./RW) [de

Corrosion behaviour of high temperature alloys in the cooling gas of high temperature reactors

International Nuclear Information System (INIS)

Quadakkers, W.J.; Schuster, H.

1989-01-01

The reactive impurities in the primary cooling helium of advanced high temperature gas cooled reactors (HTGR) can cause oxidation, carburization or decarburization of the heat exchanging metallic components. By studies of the fundamental aspects of the corrosion mechanisms it became possible to define operating conditions under which the metallic construction materials show, from the viewpoint of technical application, acceptable corrosion behaviour. By extensive test programmes with exposure times of up to 30,000 hours, a data base has been obtained which allows a reliable extrapolation of the corrosion effects up to the envisaged service lives of the heat exchanging components. (author). 6 refs, 7 figs

High-temperature gas-cooled reactors and process heat

International Nuclear Information System (INIS)

Kasten, P.R.

1980-01-01

High-Temperature Gas-Cooled Reactors (HTGRs) are fueled with ceramic-coated microspheres of uranium and thorium oxides/carbides embedded in graphite blocks which are cooled with helium. Promising areas of HTGR application are in cogeneration, energy transport using Heat Transfer Salt, recovery of oils from oil shale, steam reforming of methane for chemical production, coal gasification, and in energy transfer using chemical heat jpipes in the long term. Further, HTGRs could be used as the energy source for hydrogen production through thermochemical water splitting in the long term. The potential market for Process Heat HTGRs is 100-200 large units by about the year 2020

The cryogenic helium cooling system for the Tokamak physics experiment

International Nuclear Information System (INIS)

Felker, B.; Slack, D.S.; Wendland, C.R.

1995-01-01

The Tokamak Physics Experiment (TPX) will use supercritical helium to cool all the magnets and supply helium to the Vacuum cryopumping subsystem. The heat loads will come from the standard steady state conduction and thermal radiation sources and from the pulsed loads of the nuclear and eddy currents caused by the Central Solenoid Coils and the plasma positioning coils. The operations of the TPX will begin with pulses of up to 1000 seconds in duration every 75 minutes. The helium system utilizes a pulse load leveling scheme to buffer out the effects of the pulse load and maintain a constant cryogenic plant operation. The pulse load leveling scheme utilizes the thermal mass of liquid and gaseous helium stored in a remote dewar to absorb the pulses of the tokamak loads. The mass of the stored helium will buffer out the temperature pulses allowing 5 K helium to be delivered to the magnets throughout the length of the pulse. The temperature of the dewar will remain below 5 K with all the energy of the pulse absorbed. This paper will present the details of the heat load sources, of the pulse load leveling scheme operations, a partial helium schematic, dewar temperature as a function of time, the heat load sources as a function of time and the helium temperature as a function of length along the various components that will be cooled

Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

Energy Technology Data Exchange (ETDEWEB)

S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

2008-04-01

A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

International Nuclear Information System (INIS)

S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

2008-01-01

A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: (1) Identifies pre-conceptual design requirements; (2) Develops test loop equipment schematics and layout; (3) Identifies space allocations for each of the facility functions, as required; (4) Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems; (5) Identifies pre-conceptual utility and support system needs; and (6) Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs

The real gas behaviour of helium as a cooling medium for high-temperature reactors

International Nuclear Information System (INIS)

Hewing, G.

1977-01-01

The article describes the influence of the real gas behaviour on the variables of state for the helium gas and the effects on the design of high-temperature reactor plants. After explaining the basic equations for describing variables and changes of state of the real gas, the real and ideal gas behaviour is analysed. Finally, the influence of the real gas behaviour on the design of high-temperature reactors in one- and two-cycle plants is investigated. (orig.) [de

The Design of High Reliability Magnetic Bearing Systems for Helium Cooled Reactor Machinery

International Nuclear Information System (INIS)

Swann, M.; Davies, N.; Jayawant, R.; Leung, R.; Shultz, R.; Gao, R.; Guo, Z.

2014-01-01

The requirements for magnetic bearing equipped machinery used in high temperature, helium cooled, graphite moderated reactor applications present a set of design considerations that are unlike most other applications of magnetic bearing technology in large industrial rotating equipment, for example as used in the oil and gas or other power generation applications. In particular, the bearings are typically immersed directly in the process gas in order to take advantage of the design simplicity that comes about from the elimination of ancillary lubrication and cooling systems for bearings and seals. Such duty means that the bearings will usually see high temperatures and pressures in service and will also typically be subject to graphite particulate and attendant radioactive contamination over time. In addition, unlike most industrial applications, seismic loading events become of paramount importance for the magnetic bearings system, both for actuators and controls. The auxiliary bearing design requirements, in particular, become especially demanding when one considers that the whole mechanical structure of the magnetic bearing system is located inside an inaccessible pressure vessel that should be rarely, if ever, disassembled over the service life of the power plant. Lastly, many machinery designs for gas cooled nuclear power plants utilize vertical orientation. This circumstance presents its own unique requirements for the machinery dynamics and bearing loads. Based on the authors’ experience with machine design and supply on several helium cooled reactor projects including Ft. St. Vrain (US), GT-MHR (Russia), PBMR (South Africa), GTHTR (Japan), and most recently HTR-PM (China), this paper addresses many of the design considerations for such machinery and how the application of magnetic bearings directly affects machinery reliability and availability, operability, and maintainability. Remote inspection and diagnostics are a key focus of this paper. (author)

Evaluation, Comparison and Optimization of the Compact Recuperator for the High Temperature Gas-Cooled Reactor (HTGR) Helium Turbine System

International Nuclear Information System (INIS)

Hao Haoran; Yang Xiaoyong; Wang Jie; Ye Ping; Yu Xiaoli; Zhao Gang

2014-01-01

Helium turbine system is a promising method to covert the nuclear power generated by the High Temperature Gas Cooled Reactor (HTGR) into electricity with inherent safety, compact configuration and relative high efficiency. And the recuperator is one of the key components for the HTGR helium turbine system. It is used to recover the exhaust heat out of turbine and pass it to the helium from high pressure compressor, and hence increase the cycle’s efficiency dramatically. On the other hand, the pressure drop within the recuperator will reduce the cycle efficiency, especially on low pressure side of recuperator. It is necessary to optimize the design of recuperator to achieve better performance of HTGR helium turbine system. However, this optimization has to be performed with the restriction of the size of the pressure vessel which contains the power conversion unit. This paper firstly presents an analysis to investigate the effects of flow channel geometry, recuperator’s power and size on heat transfer and pressure drop. Then the relationship between the recuperator design and system performance is established with an analytical model, followed by the evaluations of the current recuperator designs of GT-MHR, GTHTR300 and PBMR, in which several effective technical measures to optimize the recuperator are compared. Finally it is found that the most important factors for optimizing recuperator design, i.e. the cross section dimensions and tortuosity of flow channel, which can also be extended to compact intermediate heat exchangers. It turns out that a proper optimization can increase the cycle’s efficiency by 1~2 percentage, which could also raise the economy and competitiveness of future commercial HTGR plants. (author)

Incoloy 800 stands up to radiation and corrosion in high temperature gas cooled reactors

International Nuclear Information System (INIS)

Anon.

1975-01-01

Incoloy 800 has been selected for heat exchangers in helium cooled nuclear reactor prototypes for exposure to 350 to 800 0 C helium and high temperature high purity water and steam. 304H stainless steel used in heat exchangers in original design cracked in the superheater area, bellows and tubing after static pressure tests but before exposure to steam. Residual stress, chlorides, and oxygen were deduced to have caused the failures

Medium-size high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Peinado, C.O.; Koutz, S.L.

1980-08-01

This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760 0 C (1400 0 F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics [a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant] and engineered safety features

High-temperature helium embrittlement (T>=0,45Tsub(M)) of metals

International Nuclear Information System (INIS)

Batfalsky, P.

1984-06-01

High temperature helium embrittlement, swelling and irradiation creep are the main technical problem of fusion reactor materials. The expected helium production will be very high. The helium produced by (n,α)-processes precipitates into helium bubbles because its solubility in solid metals is very low. Under continuous helium production at high temperature and stress the helium bubbles grow and lead to intergranular early failure. Solution annealed foil specimens of austenitic stainless steel AISI 316 were implanted with α-particles: 1. during creep tests at 1023 K (''in-beam'' test) 2. before the creep tests at high temperature (1023 K). The creep tests have been performed within large ranges of test parameter, e.g. applied stress, temperature, helium implantation rate and helium concentration. After the creep tests the microstructure was investigated using scanning (SEM) and transmission (TEM) electron microscopy. All the helium implanted specimens showed high temperature helium embrittlement, i.e. reduction of rupture time tsub(R) and ductility epsilonsub(R) and evidence of intergranular brittle fracture. The ''in-beam'' creep tests showed greater reduction of rupture time tsub(R) and ductility than the preimplanted creep tests. The comparison of this experimentally obtained data with various theoretical models of high temperature helium embrittlement showed that within the investigated parameter ranges the mechanism controlling the life time of the samples is probably the gas driven stable growth of the helium bubbles within the grain boundaries. (orig.)

Prestressed concrete vessels suitable for helium high temperature reactors

International Nuclear Information System (INIS)

Lockett, G.E.; Kinkead, A.N.

1967-02-01

In considering prestressed concrete vessels for use with helium cooled high temperature reactors, a number of new problems arise and projected designs involve new approaches and new solutions. These reactors, having high coolant outlet temperature from the core and relatively high power densities, can be built into compact designs which permit usefully high working pressures. Consequently, steam generators and circulating units tend to be small. Although circuit activity can be kept quite low with coated particle fuels, designs which involve entry for subsequent repair are not favoured, and coupled with the preferred aim of using fully shop fabricated units within the designs with removable steam generators which involve no tube welding inside the vessel. A particular solution uses a number of slim cylindrical assemblies housed in the wall of the pressure vessel and this vessel design concept is presented. The use of helium requires very high sealing standards and one of the important requirements is a vessel design which permits leak testing during construction, so that a repair seal can be made to any faulty part in a liner seam. Very good demountable joint seals can be made without particular difficulty and Dragon experience is used to provide solutions which are suitable for prestressed concrete vessel penetrations. The concept layout is given of a vessel meeting these requirements; the basis of design is outlined and special features of importance discussed. (author)

Preliminary study on application of Pd composite membrane in helium purification system of high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Cai Jianhua; Yang Xiaoyong; Wang Jie; Yu Suyuan

2008-01-01

Helium purification system (HPS) is the main part of the helium auxiliary system of high-temperature gas-cooled reactors (HTGR), also in fusion reactors. Some exploratory work was carried out on the application of Pd composite membrane in the separation of He and H 2 . A typical single stripper permeator with recycle (SSP) system was designed, based on the design parameters of a small scale He purification test system CIGNE in CADARACHE, CEA, France, and finite element analysis method was used to solve the model. The total length of membrane module is fixed to 0.5 m. The results show that the concentration of H 2 is found to reduce from 1 000 μL/L in feed gas to 5 μL/L in the product He (the upper limitation of HPS in HTGR). And the molar ratio of product He to feed gas is 96.18% with the optimized ratio of sweep gas to retentive gas 0. 3970. It's an exponential distribution of H 2 concentration along the membrane module. The results were also compared with the other two popular designs, two stripper in series permeator (TSSP) and continuous membrane column (CMC). (authors)

Corrosion tests of high temperature alloys in impure helium

International Nuclear Information System (INIS)

Berka, Jan; Kalivodova, Jana; Vilemova, Monika; Skoumalova, Zuzana; Brabec, Petr

2014-01-01

Czech research organizations take part several projects concerning technologies and materials for advanced gas cooled reactors, as an example international project ARCHER supported by EU within FP7, also several national projects supported by Technology Agency of the Czech Republic are solved in cooperation with industrial and research organization. Within these projects the material testing program is performed. The results presented in these paper concerning high temperature corrosion and degradation of alloys (800 H, SS 316 and P91) in helium containing minor impurities (H_2, CO, CH_4, HZO) at temperatures up to 760°C. After corrosion tests (up to 1500 hours) the specimens was investigated by several methods (gravimetry, SEM-EDX, optical microscopy, hardness and micro-hardness testing etc. (author)

Supercritical Helium Cooling of the LHC Beam Screens

CERN Document Server

Hatchadourian, E; Tavian, L

1998-01-01

The cold mass of the LHC superconducting magnets, operating in pressurised superfluid helium at 1.9 K, must be shielded from the dynamic heat loads induced by the circulating particle beams, by means of beam screens maintained at higher temperature. The beam screens are cooled between 5 and 20 K by forced flow of weakly supercritical helium, a solution which avoids two-phase flow in the long, narr ow cooling channels, but still presents a potential risk of thermohydraulic instabilities. This problem has been studied by theoretical modelling and experiments performed on a full-scale dedicated te st loop.

A charge regulating system for turbo-generator gas-cooled high-temperature reactor power stations

International Nuclear Information System (INIS)

Braytenbah, A.S.; Jaegtnes, K.O.

1975-01-01

The invention relates to a regulating system for gas-cooled high-temperature reactors power stations (helium coolant), equipped with several steam-boilers, each of which deriving heat from a corresponding cooling-gas flow circulating in the reactor, so as to feed superheated steam into a main common steam-manifold and re-superheated steam into a re-superheated hot common manifold [fr

Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

International Nuclear Information System (INIS)

Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki

2003-01-01

The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 degree sign C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 degree sign C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open

High Temperature Operational Experiences of Helium Experimental Loop

International Nuclear Information System (INIS)

Kim, Chan Soo; Hong, Sung-Deok; Kim, Eung-Seon; Kim, Min Hwan

2015-01-01

The development of high temperature components of VHTR is very important because of its higher operation temperature than that of a common light water reactor and high pressure industrial process. The development of high temperature components requires the large helium loop. Many countries have high temperature helium loops or a plan for its construction. Table 1 shows various international state-of-the-art of high temperature and high pressure gas loops. HELP performance test results show that there is no problem in operation of HELP at the very high temperature experimental condition. These experimental results also provide the basic information for very high temperature operation with bench-scale intermediate heat exchanger prototype in HELP. In the future, various heat exchanger tests will give us the experimental data for GAMMA+ validation about transient T/H behavior of the IHX prototype and the optimization of the working fluid in the intermediate loop

Modular high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shepherd, L.R.

1988-01-01

The high financial risk involved in building large nuclear power reactors has been a major factor in halting investment in new plant and in bringing further technical development to a standstill. Increased public concern about the safety of nuclear plant, particularly after Chernobyl, has contributed to this stagnation. Financial and technical risk could be reduced considerably by going to small modular units, which would make it possible to build up power station capacity in small steps. Such modular plant, based on the helium-cooled high temperature reactor (HTR), offers remarkable advantages in terms of inherent safety characteristics, partly because of the relatively small size of the individual modules but more on account of the enormous thermal capacity and high temperature margins of the graphitic reactor assemblies. Assessments indicate that, in the USA, the cost of power from the modular systems would be less than that from conventional single reactor plant, up to about 600 MW(e), and only marginally greater above that level, a margin that should be offset by the shorter time required in bringing the modular units on line to earn revenue. The modular HTR would be particularly appropriate in the UK, because of the considerable British industrial background in gas-cooled reactors, and could be a suitable replacement for Magnox. The modular reactor would be particularly suited to combined heat and power schemes and would offer great potential for the eventual development of gas turbine power conversion and the production of high-temperature process heat. (author)

High-Temperature Corrosion Behavior of Alloy 617 in Helium Environment of Very High Temperature Gas Reactor

International Nuclear Information System (INIS)

Lee, Gyeong-Geun; Jung, Sujin; Kim, Daejong; Jeong, Yong-Whan; Kim, Dong-Jin

2012-01-01

Alloy 617 is a Ni-base superalloy and a candidate material for the intermediate heat exchanger (IHX) of a very high temperature gas reactor (VHTR) which is one of the next generation nuclear reactors under development. The high operating temperature of VHTR enables various applications such as mass production of hydrogen with high energy efficiency. Alloy 617 has good creep resistance and phase stability at high temperatures in an air environment. However, it was reported that the mechanical properties decreased at a high temperature in an impure helium environment. In this study, high-temperature corrosion tests were carried out at 850°C-950°C in a helium environment containing the impurity gases H_2, CO, and CH_4, in order to examine the corrosion behavior of Alloy 617. Until 250 h, Alloy 617 specimens showed a parabolic oxidation behavior at all temperatures. The activation energy for oxidation in helium environment was 154 kJ/mol. The SEM and EDS results elucidated a Cr-rich surface oxide layer, Al-rich internal oxides and depletion of grain boundary carbides. The thickness and depths of degraded layers also showed a parabolic relationship with time. A normal grain growth was observed in the Cr-rich surface oxide layer. When corrosion tests were conducted in a pure helium environment, the oxidation was suppressed drastically. It was elucidated that minor impurity gases in the helium would have detrimental effects on the high temperature corrosion behavior of Alloy 617 for the VHTR application.

High temperature embrittlement of metals by helium

International Nuclear Information System (INIS)

Schroeder, H.

1983-01-01

The present knowledge of the influence of helium on the high temperature mechanical properties of metals to be used as structural materials in fast fission and in future fusion reactors is reviewed. A wealth of experimental data has been obtained by many different experimental techniques, on many different alloys, and on different properties. This review is mostly concentrated on the behaviour of austenitic alloys -especially austenitic stainless steels, for which the data base is by far the largest - and gives only a few examples of special bcc alloys. The effect of the helium embrittlement on the different properties - tensile, fatigue and, with special emphasis, creep - is demonstrated by representative results. A comparison between data obtained from in-pile (-beam) experiments and from post-irradiation (-implantation) experiments, respectively, is presented. Theoretical models to describe the observed phenomena are briefly outlined and some suggestions are made for future work to resolve uncertainties and differences between our experimental knowledge and theoretical understanding of high temperature helium embrittlement. (author)

Numerical analysis of performance of steam reformer of methane reforming hydrogen production system connected with high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Yin Huaqiang; Jiang Shengyao; Zhang Youjie

2007-01-01

Methane conversion rate and hydrogen output are important performance indexes of the steam reformer. The paper presents numerical analysis of performance of the reformer connected with high-temperature gas-cooled reactor HTR-10. Setting helium inlet flow rate fixed, performance of the reformer was examined with different helium inlet temperature, pressure, different process gas temperature, pressure, flow rate, and different steam to carbon ratio. As the range concerned, helium inlet temperature has remarkable influence on the performance, and helium inlet temperature, process gas temperature and pressure have little influence on the performance, and improving process gas flow rate, methane conversion rate decreases and hydrogen output increases, however improving steam to carbon ratio has reverse influence on the performance. (authors)

Development of high temperature gas cooled reactor in China

Energy Technology Data Exchange (ETDEWEB)

Guo, Wentao [Paul Scherrer Institute, Villigen (Switzerland). Dept. of Nuclear Energy and Safety; Schorer, Michael [Swiss Nuclear Forum, Olten (Switzerland)

2018-02-15

High temperature gas cooled reactor (HTGR) is one of the six Generation IV reactor types put forward by Generation IV International Forum (GIF) in 2002. This type of reactor has high outlet temperature. It uses Helium as coolant and graphite as moderator. Pebble fuel and ceramic reactor core are adopted. Inherit safety, good economy, high generating efficiency are the advantages of HTGR. According to the comprehensive evaluation from the international nuclear community, HTGR has already been given the priority to the research and development for commercial use. A demonstration project of the High Temperature Reactor-Pebble-�bed Modules (HTR-PM) in Shidao Bay nuclear power plant in China is under construction. In this paper, the development history of HTGR in China and the current situation of HTR-PM will be introduced. The experiences from China may be taken as a reference by the international nuclear community.

Adsorption purification of helium coolant of high-temperature gas-cooled reactors of carbon dioxide

International Nuclear Information System (INIS)

Varezhkin, A.V.; Zel'venskij, Ya.D.; Metlik, I.V.; Khrulev, A.A.; Fedoseenkin, A.N.

1986-01-01

A series experiments on adsorption purification of helium of CO 2 using national adsorbent under the conditions characteristic of HTGR type reactors cleanup system is performed. The experimnts have been conducted under the dynamic mode with immobile adsorbent layer (CaA zeolite) at gas flow rates from 0,02 to 0,055 m/s in the pressure range from 0,8 to 5 MPa at the temperature of 273 and 293 K. It is shown that the adsorption grows with the decrease of gas rate, i.e. with increase of contact time with adsorbent. The helium pressure, growth noticeably whereas the temperature decrease from 293 to 273 K results in adsorption 2,6 times increase. The conclusion is drawn that it is advisable drying and purification of helium of CO 2 to perform separately using different zeolites: NaA - for water. CaA - for CO 2 . Estimations of purification unit parameters are realized

Helium diffusion in nickel at high temperatures

International Nuclear Information System (INIS)

Philipps, V.

1980-09-01

Helium has been implanted at certain temperatures between 800 and 1250 0 C into single and polycrystalline Ni-samples with implantation depths between 15 and 90 μm. Simultaneously the helium reemission from the sample is measured by a mass-spectrometer. It has been shown that the time dependence of the observed reemission rate is governed by volume diffusion of the helium. Measuring this time dependence as a function of temperature the helium diffusion constant has been determined. The He-diffusion is interpreted as a interstitial diffusion hindered by thermal vacancies. Depending on the implantation depth more or less of the implanted helium remains in the sample and forms large helium bubbles. (orig./GSCH)

Potential applications of high temperature helium

International Nuclear Information System (INIS)

Schleicher, R.W. Jr.; Kennedy, A.J.

1992-09-01

This paper discusses the DOE MHTGR-SC program's recent activity to improve the economics of the MHTGR without sacrificing safety performance and two potential applications of high temperature helium, the MHTGR gas turbine plant and a process heat application for methanol production from coal

Ultralow temperature helium compressor for Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Asakura, Hiroshi

1988-01-01

Ishikawajima Harima Heavy Industries Co., Ltd. started the development of an ultralow temperature helium compressor for helium liquefaction in 1984 jointly with Japan Atomic Energy Research Institute, and has delivered the first practical machine to the Superconductive Magnet Laboratory of JAERI. For a large superconductive magnet to be used in the stable state for a fusion reactor, conventional superconductive materials (NbTi, NbTi 3 Sn, etc.) must be used, being cooled forcibly with supercritical helium. The supercritical helium which is compressed above the critical pressure of 228 kPa has a stable cooling effect since the thermal conductivity does not change due to the evaporation of liquid helium. In order to maintain the temperature of the supercritical helium below 4 K before it enters a magnet, a heat exchanger is used. The compressor that IHI has developed has the ability to reduce the vapor pressure of liquid helium from atmospheric pressure to 50.7 kPa, and can attain the temperature of 3.5 K. The specification of this single stage centrifugal compressor is: mass flow rate 25 - 64 g/s, speed 80,000 rpm, adiabatic efficiency 62 - 69 %. The structure and the performance are reported. (K.I.)

Thermal optimization of the helium-cooled power leads for the SSC

International Nuclear Information System (INIS)

Demko, J.A.; Schiesser, W.E.; Carcagno, R.; McAshan, M.; McConeghy, R.

1992-01-01

The optimum thermal design of the power leads for the Superconducting Super Collider (SSC) will minimize the amount of Carnot work (which is a combination of refrigeration and liquefaction work) required. This optimization can be accomplished by the judicious selection of lead length and diameter. Even though an optimum set of dimensions is found, the final design must satisfy other physical constraints such as maximum allowable heat leak and helium vapor mass flow rate. A set of corresponding lengths and diameters has been determined that meets these requirements for the helium vapor-cooled, spiral-fin power lead design of the SSC. Early efforts by McFee and Mallon investigated optimizing power leads for cryogenic applications with no convection cooling. Later designs utilized the boiled-off helium vapor to cool the lead. One notable design for currents up to several thousand amps is presented by Efferson based on a series of recommendations discussed by Deiness. Buyanov presents many theoretical models and design formulae but does not demonstrate an approach to thermally optimizing the design of a vapor-cooled lead. In this study, a detailed numerical thermal model of a power lead design for the SSC has been developed. It was adapted from the dynamic model developed by Schiesser. This model was used to determine the optimum dimensions that minimize the Carnot refrigeration and liquefaction work due to the leads. Since the SSC leads will be cooled by supercritical helium, the flow of vapor is regulated by a control valve. These leads include a superconducting portion at the cold end. All of the material properties in the model are functions of temperature, and for the helium are functions of pressure and temperature. No pressure drop calculations were performed as part of this analysis. The diameter that minimizes the Carnot work was determined for four different lengths at a design current of 6600 amps

Superconducting cable cooling system by helium gas at two pressures

International Nuclear Information System (INIS)

Dean, J.W.

1977-01-01

Thermally contacting, oppositely streaming, cryogenic fluid streams in the same enclosure in a closed cycle changes the fluid from a cool high pressure helium gas to a cooler reduced pressure helium gas in an expander so as to be at different temperature ranges and pressures respectively in go and return legs that are in thermal contact with each other and in thermal contact with a longitudinally extending superconducting transmission line enclosed in the same cable enclosure that insulates the line from the ambient at a temperature T 1 . By first circulating the fluid from a refrigerator at one end of the line as a cool gas at a temperature range T 2 to T 3 in the go leg, then circulating the gas through an expander at the other end of the line where the gas becomes a cooler gas at a reduced pressure and at a reduced temperature T 4 and finally by circulating the cooler gas back again to the refrigerator in a return leg at a temperature range T 4 to T 5 , while in thermal contact with the gas in the go leg, and in the same enclosure therewith for compression into a higher pressure gas at T 2 in a closed cycle, where T 2 greater than T 3 and T 5 greater than T 4 , the fluid leaves the enclosure in the go leg as a gas at its coldest point in the go leg, and the temperature distribution is such that the line temperature decreases along its length from the refrigerator due to the cooling from the gas in the return leg

High temperature tensile properties of 316 stainless steel implanted with helium

International Nuclear Information System (INIS)

Hasegawa, Akira; Yamamoto, Norikazu; Shiraishi, Haruki

1993-01-01

Helium embrittlement is one of the problems in structural materials for fusion reactors. Recently, martensitic steels have been developed which have a good resistance to high-temperature helium embrittlement, but the mechanism has not yet been clarified. In this paper, tensile behaviors of helium implanted austenitic stainless steels, which are sensitive to the helium embrittlement, were studied and compared with those of martensitic steels under the same experimental conditions, and the effect of microstructure on helium embrittlement was discussed. Helium was implanted by 300 appm at 573-623 K to miniature tensile speciments of 316 austenitic steels using a cyclotron accelerator. Solution annealed (316SA) and 20% cold worked (316CW) specimens were used. Post-implantation tensile tests were carried out at 573, 873 and 973 K. Yield stress at 573 K increased with the helium implantation in 316SA and 316CW, but the yield stress changes of 316SA at 873 and 973 K were different from that of 316CW. Black-dots were observed in the as-implanted specimen and bubbles were observed in the speciments tensile-tested at 873 and 973 K. Intergranular fracture was observed at only 973 K in both of the 316SA and 316CW specimens. Therefore, cold work did not suppress the high-temperature helium embrittlement under this experimental condition. The difference in the influence of helium on type 316 steel and 9Cr martensitic steels were discussed. Test temperature change of reduction in are showed clearly that helium embrittlement did not occur in 9Cr martensitic steels but occurred in 316 austenitic steels. Fine microstructures of 9Cr martensitic steels should suppress helium embrittlement at high temperatures. (author)

Data on test results of vessel cooling system of high temperature engineering test reactor

International Nuclear Information System (INIS)

Saikusa, Akio; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

2003-02-01

High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

Utilization of multi-purpose high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Kawada, Osamu; Onuki, Yoshiaki; Wasaoka, Takeshi.

1974-01-01

Concerning the utilization of multi-purpose high temperature gas-cooled reactors, the electric power generation with gas turbines is described: features of HTR-He gas turbine power plants; the state of development of He gas turbines; and combined cycle with gas turbines and steam turbines. The features of gas turbines concern heat dissipation into the environment and the mode of load operation. Outstanding work in the development of He gas turbines is that in Hochtemperatur Helium-Turbine Project in West Germany. The power generation with combined gas turbines and steam turbines appears to be superior to that with gas turbines alone. (Mori, K.)

Safety aspects of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)

International Nuclear Information System (INIS)

Silady, F.A.; Millunzi, A.C.

1989-08-01

The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes the basic high-temperature gas-cooled reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The qualitative top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. The MHTGR safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles has been evaluated. A broad range of challenges to core heat removal have been examined which include a loss of helium pressure and a simultaneous loss of forced cooling of the core. The challenges to control of heat generation have considered not only the failure to insert the reactivity control systems, but the withdrawal of control rods. Finally, challenges to control chemical attack of the ceramic coated fuel have been considered, including catastrophic failure of the steam generator allowing water ingress or of the pressure vessels allowing air ingress. The plant's response to these extreme challenges is not dependent on operator action and the events considered encompass conceivable operator errors. In the same vein, reliance on radionuclide retention within the full particle and on passive features to perform a few key functions to maintain the fuel within acceptable conditions also reduced susceptibility to external events, site-specific events, and to acts of sabotage and terrorism. 4 refs., 14 figs., 1 tab

Forced two phase helium cooling of large superconducting magnets

International Nuclear Information System (INIS)

Green, M.A.; Burns, W.A.; Taylor, J.D.

1979-08-01

A major problem shared by all large superconducting magnets is the cryogenic cooling system. Most large magnets are cooled by some variation of the helium bath. Helium bath cooling becomes more and more troublesome as the size of the magnet grows and as geometric constraints come into play. An alternative approach to cooling large magnet systems is the forced flow, two phase helium system. The advantages of two phase cooling in many magnet systems are shown. The design of a two phase helium system, with its control dewar, is presented. The paper discusses pressure drop of a two phase system, stability of a two phase system and the method of cool down of a two phase system. The results of experimental measurements at LBL are discussed. Included are the results of cool down and operation of superconducting solenoids

Low-temperature dynamic nuclear polarization with helium-cooled samples and nitrogen-driven magic-angle spinning.

Science.gov (United States)

Thurber, Kent; Tycko, Robert

2016-03-01

We describe novel instrumentation for low-temperature solid state nuclear magnetic resonance (NMR) with dynamic nuclear polarization (DNP) and magic-angle spinning (MAS), focusing on aspects of this instrumentation that have not been described in detail in previous publications. We characterize the performance of an extended interaction oscillator (EIO) microwave source, operating near 264 GHz with 1.5 W output power, which we use in conjunction with a quasi-optical microwave polarizing system and a MAS NMR probe that employs liquid helium for sample cooling and nitrogen gas for sample spinning. Enhancement factors for cross-polarized (13)C NMR signals in the 100-200 range are demonstrated with DNP at 25K. The dependences of signal amplitudes on sample temperature, as well as microwave power, polarization, and frequency, are presented. We show that sample temperatures below 30K can be achieved with helium consumption rates below 1.3 l/h. To illustrate potential applications of this instrumentation in structural studies of biochemical systems, we compare results from low-temperature DNP experiments on a calmodulin-binding peptide in its free and bound states. Published by Elsevier Inc.

Interdiffusion of krypton and xenon in high-pressure helium

International Nuclear Information System (INIS)

Campana, R.J.; Jensen, D.D.; Epstein, B.D.; Hudson, R.G.; Baldwin, N.L.

1980-01-01

The interdiffusion of gaseous fission products in high-pressure helium is an important factor in the control of radioactivity in gas-cooled fast breeder reactors (GCFRs). As presently conceived, GCFRs use pressure-equalized and vented fuel in which fission gases released from the solid matrix oxide fuel are transported through the fuel rod interstices and internal fission product traps to the fuel assembly vents, where they are swept away to external traps and storage. Since the predominant transport process under steady-state operating conditions is interdiffusion of gaseous fission products in helium, the diffusion properties of krypton-helium and xenon-helium couples have been measured over the range of GCFR temperature and pressure conditions ( -1 ) and expected temperature dependence to the 1.66 power (Tsup(1.66)) at lower pressures and temperatures. Additional work is in progress to measure the behaviour of the krypton-helium and xenon-helium couples in GCFR fuel rod charcoal delay traps. (author)

Corrosion behaviour of high temperature alloys in impure helium environments

International Nuclear Information System (INIS)

Shindo, Masami; Quadakkers, W.J.; Schuster, H.

1986-01-01

Corrosion tests with Ni-base high temperature alloys were carried out at 900 and 950 0 C in simulated high temperature reactor helium environments. It is shown that the carburization and decarburization behaviour is strongly affected by the Cr and Ti(Al) contents of the alloys. In carburizing environments, additions of Ti, alone or in combination with Al, significantly improve the carburization resistance. In oxidizing environment, the alloys with high Cr and Al(Ti) contents are the most resistant against decarburization. In this environment alloys with additions of Ti and Al show poor oxidation resistance. The experimental results obtained are compared with a recently developed theory describing corrosion of high temperature alloys in high temperature reactor helium environments. (orig.)

Supercritical helium cooled, cabled, superconducting hollow conductors for large high field magnets

International Nuclear Information System (INIS)

Hoenig, M.O.; Iwasa, Y.; Montgomery, D.B.; Bejan, A.

1976-01-01

Within the last two years a new concept of cabled superconducting hollow conductors has been developed which are able to recover from transient instabilities by virtue of on-going, single-phase helium cooling. It has been possible to correlate small scale experimental results with an iterative computer program. The latter has been recently upgraded to include axial as well as radial heat transfer and predict more closely the chances of recovery. Nearly 1 g/s of supercritical helium has been circulated in a closed loop using a high speed centrifugal fan and up to 10 g/s using a reciprocating single pulse bellows pump. The loop is now being adapted to a 3 m length of a tightly wound 5000 A cabled hollow conductor equipped with pulse coils designed to fit inside a water cooled Bitter magnet. The combination will allow for a steady background field of 7.5 t with a 2 t superimposed pulse. (author)

Performance of cold compressors in a cooling system of an R and D superconducting coil cooled with subcooled helium

International Nuclear Information System (INIS)

Hamaguchi, S.; Imagawa, S.; Yanagi, N.; Takahata, K.; Maekawa, R.; Mito, T.

2006-01-01

The helical coils of large helical device (LHD) have been operated in saturated helium at 4.4 K and plasma experiments have been carried out at magnetic fields lower than 3 T for 8 years. Now, it is considered that the cooling system of helical coils will be improved to enhance magnetic fields in 2006. In the improvement, the helical coils will be cooled with subcooled helium and the operating temperature of helical coils will be lowered to achieve the designed field of 3 T and enhance cryogenic stabilities. Two cold compressors will be used in the cooling system of helical coils to generate subcooled helium. In the present study, the performance of cold compressors has been investigated, using a cooling system of R and D coil, to apply cold compressors to the cooling system of helical coils. Actual surge lines of cold compressors were observed and the stable operation area was obtained. Automatic operations were also performed within the area. In the automatic operations, the suitable pressure of a saturated helium bath, calculated from the rotation speed of the 1st cold compressor, was regulated by bypass valve. From these results, stable operations will be expected in the cooling system of helical coils

Unexpected mobility of OH+ and OD+ molecular ions in cooled helium gas

International Nuclear Information System (INIS)

Isawa, R; Yamazoe, J; Tanuma, H; Ohtsuki, K

2012-01-01

Mobilities of OH + and OD + ions in cooled helium gas have been measured at gas temperature of 4.3 K. Measured mobilities of both ions as a function of an effective temperature T eff show a minimum around 80 K, and they are approaching to the polarization limits at very low T eff . These findings will be related to the extremely strong anisotropy of the interaction potential between the molecular ion and helium atom.

Test results from a helium gas-cooled porous metal heat exchanger

International Nuclear Information System (INIS)

North, M.T.; Rosenfeld, J.H.; Youchison, D.L.

1996-01-01

A helium-cooled porous metal heat exchanger was built and tested, which successfully absorbed heat fluxes exceeding all previously tested gas-cooled designs. Helium-cooled plasma-facing components are being evaluated for fusion applications. Helium is a favorable coolant for fusion devices because it is not a plasma contaminant, it is not easily activated, and it is easily removed from the device in the event of a leak. The main drawback of gas coolants is their relatively poor thermal transport properties. This limitation can be removed through use of a highly efficient heat exchanger design. A low flow resistance porous metal heat exchanger design was developed, based on the requirements for the Faraday shield for the International Thermonuclear Experimental Reactor (ITER) device. High heat flux tests were conducted on two representative test articles at the Plasma Materials Test Facility (PMTF) at Sandia National Laboratories. Absorbed heat fluxes as high as 40 MW/m 2 were successfully removed during these tests without failure of the devices. Commercial applications for electronics cooling and other high heat flux applications are being identified

The world trends of high temperature gas-cooled reactors and the mode of utilization

International Nuclear Information System (INIS)

Ishikawa, Hiroshi; Shimokawa, Jun-ichi

1974-01-01

After a long period of research and development, high temperature gas-cooled reactors are going to enter the practical stage. The combination of a HTGR with a closed cycle helium gas turbine is advantageous in thermal efficiency, reduction of environmental impact and economy. In recent years, the direct utilization of nuclear heat energy in industries has been attracting interest. The multi-purpose utilization of high temperature gas-cooled reactors is thus now the world trend. Reviewing the world developments in this field, the following matters are described: (1) development of HTGRs in the U.K., West Germany, France and the United States; (2) development of He gas turbine, etc. in West Germany; and (3) multi-purpose utilization of HTGRs in West Germany and Japan. (Mori, K.)

Evaluation of US demo helium-cooled blanket options

International Nuclear Information System (INIS)

Wong, C.P.C.; McQuillan, B.W.; Schleicher, R.W.

1995-10-01

A He-V-Li blanket design was developed as a candidate for the U.S. fusion demonstration power plant. This paper presents an 18 MPa helium-cooled, lithium breeder, V-alloy design that can be coupled to the Brayton cycle with a gross efficiency of 46%. The critical issue of designing to high gas pressure and the compatibility between helium impurities and V-alloy are addressed

Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

Science.gov (United States)

Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

2017-12-01

The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

Directory of Open Access Journals (Sweden)

Grodzki Marcin

2017-12-01

Full Text Available The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an ‘early design’ variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit. A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

Manufacturing and joining technologies for helium cooled divertors

International Nuclear Information System (INIS)

Aktaa, J.; Basuki, W.W.; Weber, T.; Norajitra, P.; Krauss, W.; Konys, J.

2014-01-01

Highlights: • The manufacturing and joining technologies developed at KIT for helium cooled divertors are reviewed and critically discussed. • Various technologies have been pursued and further developed aiming divertor components with very high quality and sufficient reliability. • Very promising routes have been found for which however still R and D works are necessary. • Technologies developed are also useful for other divertor and even blanket concepts, particularly those with tungsten armor. - Abstract: In the helium cooled (HC) divertor, developed at KIT for a fusion power plant, tungsten has been selected as armor as well as structural material due to its crucial properties: high melting point, very low sputtering yield, good thermal conductivity, high temperature strength, low thermal expansion and low activation. Thereby the armor tungsten is attached to the structural tungsten by thermally conductive joint. Due to the brittleness of tungsten at low temperatures its use as structural material is limited to the high temperature part of the component and a structural joint to the reduced activation ferritic martensitic steel EUROFER97 is foreseen. Hence, to realize the selected hybrid material concept reliable tungsten–steel and tungsten–tungsten joints have been developed and will be reported in this paper. In addition, the modular design of the HC divertor requires tungsten armor tiles and tungsten structural thimbles to be manufactured in high numbers with very high quality. Due to the high strength and low temperature brittleness of tungsten special manufacturing techniques need to be developed for the production of parts with no cavities inside and/or surface flaws. The main achievement in developing the respective manufacturing technologies will be presented and discussed. To achieve the objectives mentioned above various manufacturing and joining technologies are pursued. Their later applicability depends on the level of development

Numerical simulations of helium flow through prismatic fuel elements of very high temperature reactors

International Nuclear Information System (INIS)

Ribeiro, Felipe Lopes; Pinto, Joao Pedro C.T.A.

2013-01-01

The 4 th generation Very High Temperature Reactor (VHTR) most popular concept uses a graphite-moderated and helium cooled core with an outlet gas temperature of approximately 1000 deg C. The high output temperature allows the use of the process heat and the production of hydrogen through the thermochemical iodine-sulfur process as well as highly efficient electricity generation. There are two concepts of VHTR core: the prismatic block and the pebble bed core. The prismatic block core has two popular concepts for the fuel element: multihole and annular. In the multi-hole fuel element, prismatic graphite blocks contain cylindrical flow channels where the helium coolant flows removing heat from cylindrical fuel rods positioned in the graphite. In the other hand, the annular type fuel element has annular channels around the fuel. This paper shows the numerical evaluations of prismatic multi-hole and annular VHTR fuel elements and does a comparison between the results of these assembly reactors. In this study the analysis were performed using the CFD code ANSYS CFX 14.0. The simulations were made in 1/12 fuel element models. A numerical validation was performed through the energy balance, where the theoretical and the numerical generated heat were compared for each model. (author)

The evolution of US helium-cooled blankets

International Nuclear Information System (INIS)

Wong, C.P.C.; Schultz, K.R.; Cheng, E.T.

1991-01-01

This paper reviews and compares four helium-cooled fusion reactor blanket designs. These designs represent generic configurations of using helium to cool fusion reactor blankets that were studied over the past 20 years in the United States of America (US). These configurations are the pressurized module design, the pressurized tube design, the solid particulate and gas mixture design, and the nested shell design. Among these four designs, the nested shell design, which was invented for the ARIES study, is the simplest in configuration and has the least number of critical issues. Both metallic and ceramic-composite structural materials can be used for this design. It is believed that the nested shell design can be the most suitable blanket configuration for helium-cooled fusion power and experimental reactors. (orig.)

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

International Nuclear Information System (INIS)

Harvego, E.A.; Reza, S.M.M.; Richards, M.; Shenoy, A.

2006-01-01

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 and 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 deg. 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 deg. 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

Cooling performance of helium-gas/water coolers in HENDEL

International Nuclear Information System (INIS)

Inagaki, Yoshiyuki; Takada, Shoji; Hayashi, Haruyoshi; Kobayashi, Toshiaki; Ohta, Yukimaru; Shimomura, Hiroaki; Miyamoto, Yoshiaki

1994-01-01

The helium engineering demonstration loop (HENDEL) has four helium-gas/water coolers where the cooling water flows in the tubes and helium gas on the shell side. Their cooling performance was studied using the operational data from 1982 to 1991. The heat transfer of helium gas on the shell was obtained for segmental and step-up baffle type coolers. Also, the change with operation time was investigated. The cooling performance was lowered by the graphite powder released from the graphite components for several thousand hours and thereafter recovered because the graphite powder from the components was reduced and the powder in the cooler shell was blown off during the operation. (orig.)

The modular high-temperature gas-cooled reactor - a new production reactor

International Nuclear Information System (INIS)

Nulton, J.D.

1990-01-01

One of the reactor concepts being considered for application as a new production reactor (NPR) is a 350-MW(thermal) modular high-temperature gas-cooled reactor (MHTGR). The proposed MHTGR-NPR is based on the design of the commercial MHTGR and is being developed by a team that includes General Atomics and Combustion Engineering. The proposed design includes four modules combined into a production block that includes a shared containment, a spent-fuel storage facility, and other support facilities. The MHTGR has a helium-cooled, graphite-moderated, graphite-reflected annular core formed from prismatic graphite fuel blocks. The MHTGR fuel consists of highly enriched uranium oxycarbide (UCO) microsphere fuel particles that are coated with successive layers of pyrolytic carbon (PyC) and silicon carbide (SiC). Tritium-producing targets consist of enriched 6 Li aluminate microsphere target particles that are coated with successive layers of PyC and SiC similar to the fuel microspheres. Normal reactivity control is implemented by articulated control rods that can be inserted into channels in the inner and outer reflector blocks. Shutdown heat removal is accomplished by a single shutdown heat exchanger and electric motor-driven circulator located in the bottom of the reactor vessel. Current plans are to stack spent fuel elements in dry, helium-filled, water-cooled wells and store them for ∼1 yr before reprocessing. All phases of MHTGR fuel reprocessing have been demonstrated

Considerations on the design of a helium circulator for a high temperature modular reactor system

International Nuclear Information System (INIS)

Dumm, K.; Donaldson, J.

1988-01-01

A modular helium cooled, high temperature reactor system with a thermal output of 200 MW per reactor has been developed by the KWU group for cogeneration of electricity and process steam. The flow of the reactor coolant - Helium at 60 bars and 250/700 deg. C is maintained by one circulator per reactor. The circulator is driven by a variable speed Siemens asynchronous motor and is submerged in the helium primary system. For operational reasons high reliability and availability of the circulator is required. The operational requirements for the circulator design are presented in this paper. The actual design has been carried out in close cooperation with the designer and manufacturer of all submerged circulators operating in AGR plants in Great Britain, James Howden Co. Renfrew, Scotland. Design solutions received so far and mainly based on sufficiently proven components - such as oil bath lubricated bearing systems - will be described. Special attention will be paid on the necessary test work; especially for the prototype to confirm the lay out. (author). 9 figs

Fundamental conceptual design of the experimental multi-purpose high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shimokawa, Junichi; Yasuno, Takehiko; Yasukawa, Shigeru; Mitake, Susumu; Miyamoto, Yoshiaki

1975-06-01

The fundamental conceptual design of the experimental multi-purpose very high-temperature gas-cooled reactor (experimental VHTR of thermal output 50 MW with reactor outlet-gas temperature 1,000 0 C) has been carried out to provide the operation modes of the system consisting of the reactor and the heat-utilization system, including characteristics and performance of the components and safety of the plant system. For the heat-utilization system of the plant, heat distribution, temperature condition, cooling system constitution, and the containment facility are specified. For the operation of plant, testing capability of the reactor and controlability of the system are taken into consideration. Detail design is made of the fuel element, reactor core, reactivity control and pressure vessel, and also the heat exchanger, steam reformer, steam generator, helium circulator, helium-gas turbine, and helium-gas purification, fuel handling, and engineered safety systems. Emphasis is placed on providing the increase of the reactor outlet-gas temperature. Fuel element design is directed to the prismatic graphite blocks of hexagonal cross-section accommodating the hollow or tubular fuel pins sheathed in graphite sleeve. The reactor core is composed of 73 fuel columns in 7 stages, concerning the reference design MK-II. Orificing is made in the upper portion of core; one orifice for every 7 fuel columns. Average core power density is 2.5 watts/cm 3 . Fuel temperature is kept below 1,300 0 C in rated power. The main components, i.e. pressure vessel, reformer, gas turbine and intermediate heat exchanger are designed in detail; the IHX is of a double-shell and helically-wound tube coils, the reformer is of a byonet tube type, and the turbine-compressor unit is of an axial flow type (turbine in 6 stages and compressor in 16 stages). (auth.)

Very High Temperature Test of Alloy617 Compact Heat Exchanger in Helium Experimental Loop

Energy Technology Data Exchange (ETDEWEB)

Kim, Chan Soo; Park, Byung-Ha; Kim, Eung-Seon [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The Intermediate Heat eXchanger (IHX) is a key-challenged high temperature component which determines the efficiency and the economy of VHTR system. Heat generated in the VHTR fuel block is transferred from the VHTR to the intermediate loop through IHX. In the present, the shell-helical tube heat exchanger is generally used as IHX of the helium cooled reactor. Recently, a Printed Circuit Heat Exchanger (PCHE) is one of the candidates for the IHX in a VHTR because its operation temperature and pressure are larger than any other compact heat exchanger types. These test results show that there is no problem in operation of HELP at the very high temperature experimental condition and the alloy617 compact heat exchanger can be operated in the very high temperature condition above 850℃. In the future, the high temperature structural analysis will be studied to estimate the thermal stress during transient and thermal shock condition. The conditions and evaluation standard for the alloy 617 diffusion bonding will be minutely studied to fabricate the large-scale PCHE for the high temperature condition.

Development of THYDE-HTGR: computer code for transient thermal-hydraulics of high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Hirano, Masashi; Hada, Kazuhiko

1990-04-01

The THYDE-HTGR code has been developed for transient thermal-hydraulic analyses of high-temperature gas-cooled reactors, based on the THYDE-W code. THYDE-W is a code developed at JAERI for the simulation of Light Water Reactor plant dynamics during various types of transients including loss-of-coolant accidents. THYDE-HTGR solves the conservation equations of mass, momentum and energy for compressible gas, or single-phase or two-phase flow. The major code modification from THYDE-W is to treat helium loops as well as water loops. In parallel to this, modification has been made for the neutron kinetics to be applicable to helium-cooled graphite-moderated reactors, for the heat transfer models to be applicable to various types of heat exchangers, and so forth. In order to assess the validity of the modifications, analyses of some of the experiments conducted at the High Temperature Test Loop of ERANS have been performed. In this report, the models applied in THYDE-HTGR are described focusing on the present modifications and the results from the assessment calculations are presented. (author)

Research and development of a high-temperature helium-leak detection system (joint research). Part 1 survey on leakage events and current leak detection technology

Energy Technology Data Exchange (ETDEWEB)

Sakaba, Nariaki; Nakazawa, Toshio; Kawasaki, Kozo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Urakami, Masao; Saisyu, Sadanori [Japan Atomic Power Co., Tokyo (Japan)

2003-03-01

In High Temperature Gas-cooled Reactors (HTGR), the detection of leakage of helium at an early stage is very important for the safety and stability of operations. Since helium is a colourless gas, it is generally difficult to identify the location and the amount of leakage when very little leakage has occurred. The purpose of this R and D is to develop a helium leak detection system for the high temperature environment appropriate to the HTGR. As the first step in the development, this paper describes the result of surveying leakage events at nuclear facilities inside and outside Japan and current gas leakage detection technology to adapt optical-fibre detection technology to HTGRs. (author)

Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

Energy Technology Data Exchange (ETDEWEB)

Middleton, Bobby [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pasch, James Jay [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kruizenga, Alan Michael [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Walker, Matthew [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

2016-01-01

This report outlines the thermodynamics of a supercritical carbon dioxide (sCO₂) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO₂ code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related to both Helium and to sCO₂ , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO₂> Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO₂ Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation

Thermal optimization of the helium-cooled power leads for the SSC

International Nuclear Information System (INIS)

Demko, J.A.; Schiesser, W.E.; Carcagno, R.; McAshan, M.; McConeghy, R.

1992-03-01

The optimum thermal design of the power leads for the Superconducting Super Collider (SSC) will minimize the amount of Carnot work (which is a combination of refrigeration and liquefaction work) required. This optimization can be accomplished by the judicious selection of lead length and diameter. Even though an optimum set of dimensions is found, the final design must satisfy other physical constraints such as maximum allowable heat leak and helium vapor mass flow rate. A set of corresponding lengths and diameters has been determined that meets these requirements for the helium vapor-cooled, spiral-fin power lead design of the SSC. Early efforts by McFee and Mallon investigated optimizing power leads for cryogenic applications with no convection cooling. Later designs utilized the boiled-off helium vapor to cool the lead. One notable design for currents up to several thousand amps is presented by Efferson based on a series of recommendations discussed by Deiness. Buyanov presents many theoretical models and design formulate but does not demonstrate an approach to thermally optimizing the design of a vapor-cooled lead. A method for optimizing superconducting magnet current leads is described by Maehata et al. The approach assumes that the helium boil-off caused by heat conduction along with power lead into the low-temperature helium is used to cool the lead. The optimum solution is found when the heat flow at the cold end is minimized.. In this study, a detailed numerical thermal model of a power lead design for the SSC has been developed. It was adapted from the dynamic model developed by Schiesser. This model was used to determine the optimum dimensions that minimize the Carnot refrigeration and liquefaction work due to the leads

Study of Transient Heat Transport Mechanisms in Superfluid Helium Cooled Rutherford-Cables

CERN Document Server

AUTHOR|(CDS)2100615

The Large Hadron Collider leverages superconducting magnets to focus the particle beam or keep it in its circular track. These superconducting magnets are composed of NbTi-cables with a special insulation that allows superfluid helium to enter and cool the superconducting cable. Loss mechanisms, e.g. continuous random loss of particles escaping the collimation system heating up the magnets. Hence, a local temperature increase can occur and lead to a quench of the magnets when the superconductor warms up above the critical temperature. A detailed knowledge about the temperature increases in the superconducting cable (Rutherford type) ensures a secure operation of the LHC. A sample of the Rutherford cable has been instrumented with temperature sensors. Experiments with this sample have been performed within this study to investigate the cooling performance of the helium in the cable due to heat deposition. The experiment uses a superconducting coil, placed in a cryostat, to couple with the magnetic field loss m...

Helium release rates and ODH calculations from RHIC magnet cooling line failure

Energy Technology Data Exchange (ETDEWEB)

Liaw, C.J.; Than, Y.; Tuozzolo, J.

2011-03-28

A catastrophic failure of the magnet cooling lines, similar to the LHC superconducting bus failure incident, could discharge cold helium into the RHIC tunnel and cause an Oxygen Deficiency Hazard (ODH) problem. A SINDA/FLUINT{reg_sign} model, which simulated the 4.5K/4 atm helium flowing through the magnet cooling system distribution lines, then through a line break into the insulating vacuum volumes and discharging via the reliefs into the RHIC tunnel, had been developed. Arc flash energy deposition and heat load from the ambient temperature cryostat surfaces are included in the simulations. Three typical areas: the sextant arc, the Triplet/DX/D0 magnets, and the injection area, had been analyzed. Results, including helium discharge rates, helium inventory loss, and the resulting oxygen concentration in the RHIC tunnel area, are reported. Good agreement had been achieved when comparing the simulation results, a RHIC sector depressurization test measurement, and some simple analytical calculations.

High-power frequency-stabilized laser for laser cooling of metastable helium at 389 nm

NARCIS (Netherlands)

Koelemeij, J.C.J.; Hogervorst, W.; Vassen, W.

2005-01-01

A high-power, frequency-stabilized laser for cooling of metastable helium atoms using the 2 S13 →3 P23 transition at 389 nm has been developed. The 389 nm light is generated by frequency doubling of a titanium:sapphire laser in an external enhancement cavity containing a lithium-triborate nonlinear

Numerical examination of temperature control in helium-cooled high flux test module of IFMIF

International Nuclear Information System (INIS)

Ebara, Shinji; Yokomine, Takehiko; Shimizu, Akihiko

2007-01-01

For long term irradiation of the International Fusion Materials Irradiation Facility (IFMIF), test specimens are needed to retain constant temperature to avoid change of its irradiation characteristics. The constant temperatures control is one of the most challenging issues for the IFMIF test facilities. We have proposed a new concept of test module which is capable of precisely measuring temperature, keeping uniform temperature with enhanced cooling performance. In the system according to the new design, cooling performances and temperature distributions of specimens were examined numerically under diverse conditions. Some transient behaviors corresponding to the prescribed temperature control mode were perseveringly simulated. It was confirmed that the thermal characteristics of the new design satisfied the severe requirement of IFMIF

Research on dynamics and experiments about auxiliary bearings for the helium circulator of the 10 MW high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Zhao, Yulan; Yang, Guojun; Liu, Xingnan; Shi, Zhengang; Zhao, Lei

2016-01-01

Highlights: • The research in this paper is based on the AMB helium circulator of HTR-10. • The dynamic rotor performance is analyzed by processing experimental data. • The mechanical bearing without lubrication can be applied in the HTR-10 system. - Abstract: The 10 MW high-temperature gas-cooled reactor (HTR-10) was constructed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University. The auxiliary bearing is utilized in this system to meet particular requirements for the reactor. The main role of the auxiliary bearing is to constrain rotor displacements and also to support the rotor when the rotor drops down, which is caused by the active magnetic bearing (AMB) failure. The auxiliary bearing needs to endure huge impact, rapid angular acceleration and thermal shock. On the one hand, complex geometrical constructions and forces applied on the system bring difficulties and restrictions to establish an appropriate model to reveal the actual dynamic process. On the other hand, large volumes of data obtained from experiments show velocities and displacements of the rotor during the rotor drop process and then can indicate the actual dynamic interactions to a great extent. The research in this paper is based on the test rig of the AMB helium circulator of HTR-10. This paper aims to analyze the dynamic performance and contact forces of the rotor by processing experimental data. A measurement to estimate forces developed due to impacts of the rotor and the auxiliary bearings is presented. It is of great significance and provides certain foundation to elaborate the rotor drop process for the AMB helium circulator of HTR-10.

High temperature friction and seizure in gas cooled nuclear reactors

International Nuclear Information System (INIS)

Cousseran, P.; Febvre, A.; Martin, R.; Roche, R.

1978-01-01

One of the most delicate problems encountered in the gas cooled nuclear reactors is the friction without lubrication in a dry and hot (800 0 C /1472 0 F) helium atmosphere even at very small velocity. The research and development programs are described together with special tribometers that operate at mode than 1000 0 C (1832 0 F) in dry helium. The most interesting test conditions and results are given for gas nitrited steels and for strongly alloyed Ni-Cr steels coated with chromium carbide by plasma sprayed. The effects of parameters live velocity, travelled distance, contact pressure, roughness, temperature and prolonged stops under charge are described together with the effects of negative phenomena like attachment and chattering [fr

What happens to liquid helium 3 at very low temperatures

International Nuclear Information System (INIS)

Dobbs, E.R.; Bedford Coll., London

1976-01-01

Liquid helium 3 forms a highly compressible, quantum fluid of fermions at its boiling point of 3.2 K. As it is cooled this fluid condenses into a viscous Fermi liquid until at temperatures of a few millikelvin it is transformed into one of two superfluids, A or B. The A phase is an anisotoropic superfluid of the Anderson-Brinkman-Morel type, while the B phase is an isotropic superfluid of the Balian-Werthamer type. The properties of these superfluids will be discussed and contrasted with those of superfluid helium 4 and metallic superconductors. (orig.) [de

Accident tolerant high-pressure helium injection system concept for light water reactors

International Nuclear Information System (INIS)

Massey, Caleb; Miller, James; Vasudevamurthy, Gokul

2016-01-01

Highlights: • Potential helium injection strategy is proposed for LWR accident scenarios. • Multiple injection sites are proposed for current LWR designs. • Proof-of-concept experimentation illustrates potential helium injection benefits. • Computational studies show an increase in pressure vessel blowdown time. • Current LOCA codes have the capability to include helium for feasibility calculations. - Abstract: While the design of advanced accident-tolerant fuels and structural materials continues to remain the primary focus of much research and development pertaining to the integrity of nuclear systems, there is a need for a more immediate, simple, and practical improvement in the severe accident response of current emergency core cooling systems. Current blowdown and reflood methodologies under accident conditions still allow peak cladding temperatures to approach design limits and detrimentally affect the integrity of core components. A high-pressure helium injection concept is presented to enhance accident tolerance by increasing operator response time while maintaining lower peak cladding temperatures under design basis and beyond design basis scenarios. Multiple injection sites are proposed that can be adapted to current light water reactor designs to minimize the need for new infrastructure, and concept feasibility has been investigated through a combination of proof-of-concept experimentation and computational modeling. Proof-of-concept experiments show promising cooling potential using a high-pressure helium injection concept, while the developed choked-flow model shows core depressurization changes with added helium injection. Though the high-pressure helium injection concept shows promise, future research into the evaluation of system feasibility and economics are needed.Classification: L. Safety and risk analysis

Conceptual design of a forced-flow-cooled 20-kA current lead using Ag-alloy-sheathed Bi-2223 high-temperature superconductors

International Nuclear Information System (INIS)

Heller, R.

1994-11-01

The use of high-temperature superconductors in current leads to reduce refrigeration power has been investigated by many groups in the past. Most used YBCO and Bi-2212 bulk superconductors, although their critical current density is not very high. In this paper, BI-2223 HTSC tapes sheathed with Ag alloys are used in the design of a 20-kA current lead because of their higher critical current in medium magnetic fields. The lead current of 20 kA is related to the coil current of the planned stellarator WENDELSTEIN 7-X. Forced-now helium cooling has been used in the design, allowing position-independent and well-controlled operation. The design characteristics of the lead are presented and 4-K helium cooling of the whole lead, as well as 60-K helium cooling of the copper part of the lead, is discussed. The power consumption at zero current, and the lead's behaviour in case of loss of coolant flow, are given, The results of the design allow extrapolation to current leads of the 50-kA range

'In-beam' simulation of high temperature helium embrittlement of DIN 1.4970 austenitic stainless steel

International Nuclear Information System (INIS)

Schroeder, H.; Batfalsky, P.

1982-01-01

This work describes a facility for high temperature creep rupture tests during homogeneous helium implantation. This 'in-beam' creep testing facility is used to simulate helium embrittlement effects which will be very important for first wall materials of future fusion reactors operated at high temperatures. First results for DIN 1.4970 austenitic stainless steel clearly demonstrate differences between samples 'in-beam' tested at 1073 K and those creep tested at the same temperature after room temperature helium implantation. The specimens ruptured 'in-beam' have much shorter lifetimes and lower ductility than the specimens tested after room temperature implantation. There are also differences in the microstructures, concerning helium bubble sizes and densities in matrix and grain boundaries. These microstructural differences may be a key for the understanding of the more severe helium embrittlement effects 'in-beam' as compared to creep tests performed after room temperature implantation. (orig.)

Hypothetical accident scenario analyses for a 250-MW(t) modular high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

1985-11-01

This paper describes calculations performed to characterize the inherent safety of a 250-MW(t), 100-MW(e), pebble bed modular high temperature gas-cooled reactor (HTGR) design with vertical in-line arrangement (i.e., upflow core with steam generators directly above the core). A variety of postulated accident sequences involving combinations of loss of forced primary coolant (helium) circulation, loss of primary coolant pressurization, and loss of heat sink were studied and were discussed

The modular high-temperature gas-cooled reactor (MHTGR) in the US

International Nuclear Information System (INIS)

Neylan, A.J.; Graf, D.F.; Millunzi, A.C.

1987-01-01

GA Technologies Inc. and other U.S. corporations, in a cooperative program with the U.S. Department of Energy, is developing a Modular High-Temperature Gas-Cooled Reactor (MHTGR) that will provide highly reliable, economic, nuclear power. The MHTGR system assures maximum safety to the public, the owner/operator, and the environment. The MHTGR is being designed to meet and exceed rigorous requirements established by the user industry for availability, operation and maintenance, plant investment protection, safety and licensing, siting flexibility and economics. The plant will be equally attractive for deployment and operation in the U.S., other major industrialized nations including Korea, Japan, and the Republic of China, as well as the developing nations. The High-Temperature Gas-Cooled Reactor (HTGR) is an advanced, third generation nuclear power system which incorporates distinctive technical features, including the use of pressurized helium as a coolant, graphite as the moderator and core structural material, and fuel in the form of ceramic coated uranium particles. The modular HTGR builds upon generic gas-cooled reactor experience and specific HTGR programs and projects. The MHTGR offers unique technological features and the opportunity for the cooperative international development of an advanced energy system that will help assure adaquate world energy resources for the future. Such international joint venturing of energy development can offer significant benefits to participating industries and governments and also provides a long term solution to the complex problems of the international balance of payments

Design, fabrication, and testing of a helium-cooled module for the ITER divertor

International Nuclear Information System (INIS)

Baxi, C.B.; Smith, J.P.; Youchison, D.

1994-08-01

The International Thermonuclear Reactor (ITER) will have a single-null divertor with total power flow of 200 MW and a peak heat flux of about 5 MW/m 2 . The reference coolant for the divertor is water. However, helium is a viable alternative and offers advantages from safety considerations, such as excellent radiation stability and chemical inertness. In order to prove the feasibility of helium cooling at ITER relevant heat flux conditions, General Atomics designed, fabricated, and tested a helium-cooled divertor module. The module was made from dispersion strengthened copper, with a heat flux surface 25 mm wide and 80 mm long, designed for twice the ITER divertor heat flux. Different techniques were examined to enhance the heat transfer, which in turn reduced the flow and pumping power required to cool the module. It was concluded that an extended surface was the most practical solution. An optimization study was performed to find the best extended surface parameters. The optimum extended surface geometry consisted of fins: 10 mm high, 0.4 mm thick with a 1 mm pitch. It was estimated to require a pumping power of 150 W to remove 20 kW of power. This is more than an order of magnitude reduction in pumping power requirement, compared to smooth surface. The module was fabricated by electric discharge machining (EDM) process. The testing was carried out at SNLA during August 1993. The testing confirmed the design calculations. The peak heat flux during the test was 10 MW/m 2 applied over a surface area of 20 cm 2 . The pumping power calculated from flow rate and pressure drop measurement was about 160 W, which was less than 1% of the power removed. It is planned to test the module to higher temperature limits and higher heat fluxes during coming months. As a result of this effort we conclude that helium cooling of the ITER divertor is feasible without requiring a very large helium pressure or a large pumping power

Parametric studies on different gas turbine cycles for a high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Wang Jie; Gu Yihua

2005-01-01

The high temperature gas-cooled reactor (HTGR) coupled with turbine cycle is considered as one of the leading candidates for future nuclear power plants. In this paper, the various types of HTGR gas turbine cycles are concluded as three typical cycles of direct cycle, closed indirect cycle and open indirect cycle. Furthermore they are theoretically converted to three Brayton cycles of helium, nitrogen and air. Those three types of Brayton cycles are thermodynamically analyzed and optimized. The results show that the variety of gas affects the cycle pressure ratio more significantly than other cycle parameters, however, the optimized cycle efficiencies of the three Brayton cycles are almost the same. In addition, the turbomachines which are required for the three optimized Brayton cycles are aerodynamically analyzed and compared and their fundamental characteristics are obtained. Helium turbocompressor has lower stage pressure ratio and more stage number than those for nitrogen and air machines, while helium and nitrogen turbocompressors have shorter blade length than that for air machine

Cryogenic thermometer calibration system using a helium cooling loop and a temperature controller [for LHC magnets

CERN Document Server

Chanzy, E; Thermeau, J P; Bühler, S; Joly, C; Casas-Cubillos, J; Balle, C

1998-01-01

The IPN-Orsay and CERN are designing in close collaboration a fully automated cryogenic thermometer calibration facility which will calibrate in 3 years 10,000 cryogenic thermometers required for the Large Hadron Collider (LHC) operation. A reduced-scale model of the calibration facility has been developed, which enables the calibration of ten thermometers by comparison with two rhodium-iron standard thermometers in the 1.8 K to 300 K temperature range under vacuum conditions. The particular design, based on a helium cooling loop and an electrical temperature controller, gives good dynamic performances. This paper describes the experimental set-up and the data acquisition system. Results of experimental runs are also presented along with the estimated global accuracy for the calibration. (3 refs).

High temperature indentation of helium-implanted tungsten

Energy Technology Data Exchange (ETDEWEB)

Gibson, James S.K.-L., E-mail: james.gibson@materials.ox.ac.uk [Oxford University, Department of Materials, Parks Road, Oxford OX1 3PH (United Kingdom); Roberts, Steve G. [Oxford University, Department of Materials, Parks Road, Oxford OX1 3PH (United Kingdom); Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Armstrong, David E.J. [Oxford University, Department of Materials, Parks Road, Oxford OX1 3PH (United Kingdom)

2015-02-11

Nanoindentation has been performed on tungsten, unimplanted and helium-implanted to ~600 appm, at temperatures up to 750 °C. The hardening effect of the damage was 0.90 GPa at 50 °C, but is negligible above 450 °C. The hardness value at a given temperature did not change on re-testing after heating to 750 °C. This suggests that the helium is trapped in small vacancy complexes that are stable to at least 750 °C, but which can be bypassed due to increased dislocation mobility (cross slip or climb) above 450 °C.

Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

Science.gov (United States)

Larson, V. R.; Gunn, S. V.; Lee, J. C.

1975-01-01

The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

Energy Technology Data Exchange (ETDEWEB)

Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra; Su’ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Aji, Indarta K. [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

2016-03-11

High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

Advanced CSiC composites for high-temperature nuclear heat transport with helium, molten salts, and sulphur-iodine thermochemical hydrogen process fluids

International Nuclear Information System (INIS)

Peterson, P.F.; Forsberg, Ch.W.; Pickard, P.S.

2004-01-01

This paper discusses the use of liquid-silicon-impregnated (LSI) carbon-carbon composites for the development of compact and inexpensive heat exchangers, piping, vessels and pumps capable of operating in the temperature range of 800 to 1 100 deg C with high-pressure helium, molten fluoride salts, and process fluids for sulfur-iodine thermochemical hydrogen production. LSI composites have several potentially attractive features, including ability to maintain nearly full mechanical strength to temperatures approaching 1 400 deg C, inexpensive and commercially available fabrication materials, and the capability for simple forming, machining and joining of carbon-carbon performs, which permits the fabrication of highly complex component geometries. In the near term, these materials may prove to be attractive for use with a molten-salt intermediate loop for the demonstration of hydrogen production with a gas-cooled high temperature reactor. In the longer term, these materials could be attractive for use with the molten-salt cooled advanced high temperature reactor, molten salt reactors, and fusion power plants. (author)

Corrosion of nickel-base heat resistant alloys in simulated VHTR coolant helium at very high temperatures

International Nuclear Information System (INIS)

Shindo, Masami; Kondo, Tatsuo

1976-01-01

A comparative evaluation was made on three commercial nickel-base heat resistant alloys exposed to helium-base atmosphere at 1000 0 C, which contained several impurities in simulating the helium cooled very high temperature nuclear reactor (VHTR) environment. The choice of alloys was made so that the effect of elements commonly found in commercial alloys were typically examined. The corrosion in helium at 1000 0 C was characterized by the sharp selection of thermodynamically unstable elements in the oxidizing process and the resultant intergranular penetration and internal oxidation. Ni-Cr-Mo-W type solution hardened alloy such as Hastelloy-X showed comparatively good resistance. The alloy containing Al and Ti such as Inconel-617 suffered adverse effect in contrast to its good resistance to air oxidation. The alloy nominally composed only of noble elements, Ni, Fe and Mo, such as Hastelloy-B showed least apparent corrosion, while suffered internal oxidation due to small amount of active impurities commonly existing in commercial heats. The results were discussed in terms of selection and improvement of alloys for uses in VHTR and the similar systems. (auth.)

Material development for gas-cooled high temperature reactors for the production of nuclear process heat

International Nuclear Information System (INIS)

Nickel, H.

1977-04-01

In the framework of the material development for gas-cooled high temperature reactors, considerable investigations of the materials for the reactor core and the primary cicuit are being conducted. Concerning the core components, the current state-of-the-art and the objectives of the development work on the spherical fuel elements, coated particles and structural graphite are discussed. As an example of the structural graphite, the non-replaceable reflector of the process heat reactor is discussed. The primary circuit will be constructed mainly from metallic materials, although some ceramics are also being considered. Components of interest are hot gas ducts, liners, methane reformer tubes and helium-helium intermediate heat exchangers. The gaseous impurities present in the helium coolant may cause oxidation and carburization of the nickel-base and iron-base alloys envisaged for use in these components, with a possible associated adverse effect on the mechanical properties such as creep and fatigue. Test capacity has therefore been installed to investigate materials behaviour in simulated reactor helium under both constant and alternating stress conditions. The first results on the creep behaviour of several alloys in impure helium are presented and discussed. (orig./GSC) [de

Operating experience using venturi flow meters at liquid helium temperature

International Nuclear Information System (INIS)

Wu, K.C.

1992-01-01

Experiences using commercial venturi to measure single phase helium flow near 4 K (degree Kelvin) for cooling superconducting magnets have been presented. The mass flow rate was calculated from the differential pressure and the helium density evaluated from measured pressure and temperature. The venturi flow meter, with a full range of 290 g/s (0.29 Kg/s) at design conditions, has been found to be reliable and accurate. The flow measurements have been used, with great success, for evaluating the performance of a cold centrifugal compressor, the thermal acoustic heat load of a cryogenic system and the cooling of a superconducting magnet after quench

Reduction of circulation power for helium-cooled fusion reactor blanket using additive CO{sub 2} gas

Energy Technology Data Exchange (ETDEWEB)

Lee, Yeon-Gun [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Department of Nuclear and Energy Engineering, Jeju National University, 102 Jejudaehakno, Jeju-si 690-756, Jeju (Korea, Republic of); Park, Il-Woong [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Lee, Dong Won [Nuclear Fusion Engineering Development Center, Korea Atomic Energy Research Institute, Daedeokdaero 989 beon-gil, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Kim, Eung-Soo, E-mail: kes7741@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of)

2015-11-15

Helium (He) cooling requires large circulation power to remove high heat from plasma side and nuclear heating by high energy neutron in fusion reactors due to its low density. Based on the recent findings that the heat transfer capability of the light gas can be enhanced by mixing another heavier gas, this study adds CO{sub 2} to a reference helium coolant and evaluates the cooling performance of the binary mixture for various compositions. To assess the cooling performance, computational fluid dynamic (CFD) analyses on the KO HCML (Korea Helium Cooled Molten Lithium) TBM are conducted. As a result, it is revealed that the binary mixing of helium, which has favorable thermophysical properties but the density, with a heavier noble gas or an unreactive gas significantly reduces the required circulation power by an order of magnitude with meeting the thermal design requirements. This is attributed to the fact that the density can be highly increased with small amount of a heavier gas while other gas properties are kept relatively comparable. The optimal CO{sub 2} mole fraction is estimated to be 0.4 and the circulation power, in this case, can be reduced to 13% of that of pure helium. This implies that the thermal efficiency of a He-cooled blanket system can be fairly enhanced by means of the proposed binary mixing.

A helium-cooled blanket design of the low aspect ratio reactor

International Nuclear Information System (INIS)

Wong, C.P.; Baxi, C.B.; Reis, E.E.; Cerbone, R.; Cheng, E.T.

1998-03-01

An aggressive low aspect ratio scoping fusion reactor design indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of > 4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 436 C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m(2), respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650 C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh

R and D programme on generation IV nuclear energy systems: the high temperatures gas-cooled reactors

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Billot, P.; Anzieu, P.; Brossard, P.

2005-01-01

The Generation IV Technology Roadmap selected, among others, a sequenced development of advanced high temperature gas cooled reactors as one of the main focus for R and D on future nuclear energy systems. The selection of this research objective originates both from the significance of high temperature and fast neutrons for nuclear energy to meet the needs for a sustainable development for the medium-long term (2020/2030 and beyond), and from the significant common R and D pathway that supports both medium term industrial projects and more advanced versions of gas cooled reactors. The first step of the 'Gas Technology Path' aims to support the development of a modular HTR to meet specific international market needs around 2020. The second step is a Very High Temperature Reactor - VHTR (>950 C) - to efficiently produce hydrogen through thermo-chemical or electro-chemical water splitting or to generate electricity with an efficiency above 50%, among other applications of high temperature nuclear heat. The third step of the Path is a Gas Fast Reactor - GFR - that features a fast-spectrum helium-cooled reactor and closed fuel cycle, with a direct or indirect thermodynamic cycle for electricity production and full recycle of actinides. Hydrogen production is also considered for the GFR. The paper succinctly presents the R and D program currently under definition and partially launched within the Generation IV International Forum on this consistent set of advanced gas cooled nuclear systems. (orig.)

Pressurized helium II-cooled magnet test facility

International Nuclear Information System (INIS)

Warren, R.P.; Lambertson, G.R.; Gilbert, W.S.; Meuser, R.B.; Caspi, S.; Schafer, R.V.

1980-06-01

A facility for testing superconducting magnets in a pressurized bath of helium II has been constructed and operated. The cryostat accepts magnets up to 0.32 m diameter and 1.32 m length with current to 3000 A. In initial tests, the volume of helium II surrounding the superconducting magnet was 90 liters. Minimum temperature reached was 1.7 K at which point the pumping system was throttled to maintain steady temperature. Helium II reservoir temperatures were easily controlled as long as the temperature upstream of the JT valve remained above T lambda; at lower temperatures control became difficult. Positive control of the temperature difference between the liquid and cold sink by means of an internal heat source appears necessary to avoid this problem. The epoxy-sealed vessel closures, with which we have had considerable experience with normal helium vacuum, also worked well in the helium II/vacuum environment

Effect of wall thickness and helium cooling channels on duct magnetohydrodynamic flows

International Nuclear Information System (INIS)

He, Qingyun; Feng, Jingchao; Chen, Hongli

2016-01-01

Highlights: • MHD flows in ducts of different wall thickness compared with wall uniform. • Study of velocity, pressure distribution in ducts MHD flows with single pass of helium cooling channels. • Comparison of three types of dual helium cooling channels and acquisition of an option for minimum pressure drop. • A single short duct MHD flow in blanket without FCI has been simulated for pressure gradient analysis. - Abstract: The concept of dual coolant liquid metal (LM) blanket has been proposed in different countries to demonstrate the technical feasibility of DEMO reactor. In the system, helium gas and PbLi eutectic, separated by structure grid, are used to cool main structure materials and to be self-cooled, respectively. The non-uniform wall thickness of structure materials gives rise to wall non-homogeneous conductance ratio. It will lead to electric current distribution changes, resulting in significant changes in the velocity distribution and pressure drop of magnetohydrodynamic (MHD) flows. In order to investigate the effect of helium channels on MHD flows, different methods of numerical simulations cases are carried out including the cases of different wall thicknesses, single pass of helium cooling channels, and three types of dual helium cooling channels. The results showed that helium tubes are able to affect the velocity distribution in the boundary layer by forming wave sharp which transfers from Hartmann boundary layer to the core area. In addition, the potential profile and pressure drop in the cases have been compared to these in the case of walls without cooling channel, and the pressure gradient of a simplified single short duct MHD flow in blanket shows small waver along the central axis in the helium channel position.

Effect of wall thickness and helium cooling channels on duct magnetohydrodynamic flows

Energy Technology Data Exchange (ETDEWEB)

He, Qingyun; Feng, Jingchao; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn

2016-02-15

Highlights: • MHD flows in ducts of different wall thickness compared with wall uniform. • Study of velocity, pressure distribution in ducts MHD flows with single pass of helium cooling channels. • Comparison of three types of dual helium cooling channels and acquisition of an option for minimum pressure drop. • A single short duct MHD flow in blanket without FCI has been simulated for pressure gradient analysis. - Abstract: The concept of dual coolant liquid metal (LM) blanket has been proposed in different countries to demonstrate the technical feasibility of DEMO reactor. In the system, helium gas and PbLi eutectic, separated by structure grid, are used to cool main structure materials and to be self-cooled, respectively. The non-uniform wall thickness of structure materials gives rise to wall non-homogeneous conductance ratio. It will lead to electric current distribution changes, resulting in significant changes in the velocity distribution and pressure drop of magnetohydrodynamic (MHD) flows. In order to investigate the effect of helium channels on MHD flows, different methods of numerical simulations cases are carried out including the cases of different wall thicknesses, single pass of helium cooling channels, and three types of dual helium cooling channels. The results showed that helium tubes are able to affect the velocity distribution in the boundary layer by forming wave sharp which transfers from Hartmann boundary layer to the core area. In addition, the potential profile and pressure drop in the cases have been compared to these in the case of walls without cooling channel, and the pressure gradient of a simplified single short duct MHD flow in blanket shows small waver along the central axis in the helium channel position.

Investigation of the loss of forced cooling test by using the high temperature engineering test reactor (HTTR) (Contract research)

International Nuclear Information System (INIS)

Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Inaba, Yoshitomo; Goto, Minoru; Tochio, Daisuke

2007-09-01

The three gas circulators trip test and the vessel cooling system stop test as the safety demonstration test by using the High Temperature engineering Test Reactor (HTTR) are under planning to demonstrate inherent safety features of High Temperature Gas-cooled Reactor. All three gas circulators to circulate the helium gas as the coolant are stopped to simulate the loss of forced cooling in the three gas circulators trip test. The stop of the vessel cooling system located outside the reactor pressure vessel to remove the residual heat of the reactor core follows the stop of all three gas circulators in the vessel cooling system stop test. The analysis of the reactor transient for such tests and abnormal events postulated during the test was performed. From the result of analysis, it was confirmed that the three gas circulators trip test and the vessel cooling system stop test can be performed within the region of the normal operation in the HTTR and the safety of the reactor facility is ensured even if the abnormal events would occur. (author)

High temperature superconductor current leads

International Nuclear Information System (INIS)

Zeimetz, B.; Liu, H.K.; Dou, S.X.

1996-01-01

Full text: The use of superconductors in high electrical current applications (magnets, transformers, generators etc.) usually requires cooling with liquid Helium, which is very expensive. The superconductor itself produces no heat, and the design of Helium dewars is very advanced. Therefore most of the heat loss, i.e. Helium consumption, comes from the current lead which connects the superconductor with its power source at room temperature. The current lead usually consists of a pair of thick copper wires. The discovery of the High Temperature Superconductors makes it possible to replace a part of the copper with superconducting material. This drastically reduces the heat losses because a) the superconductor generates no resistive heat and b) it is a very poor thermal conductor compared with the copper. In this work silver-sheathed superconducting tapes are used as current lead components. The work comprises both the production of the tapes and the overall design of the leads, in order to a) maximize the current capacity ('critical current') of the superconductor, b) minimize the thermal conductivity of the silver clad, and c) optimize the cooling conditions

Solutions for Liquid Nitrogen Pre-Cooling in Helium Refrigeration Cycles

CERN Document Server

Wagner, U

2000-01-01

Pre-cooling of helium by means of liquid nitrogen is the oldest and one of the most common process features used in helium liquefiers and refrigerators. Its two principle tasks are to allow or increase the rate of pure liquefaction, and to permit the initial cool-down of large masses to about 80 K. Several arrangements for the pre-cooling process are possible depending on the desired application. Each arrangement has its proper advantages and drawbacks. The aim of this paper is to review the possible process solutions for liquid nitrogen pre-cooling and their particularities.

Liquid helium-cooled MOSFET preamplifier for use with astronomical bolometer

Science.gov (United States)

Goebel, J. H.

1977-01-01

A liquid helium-cooled p-channel enhancement mode MOSFET, the 3N167, is found to have sufficiently low noise for use as a preamplifier with helium-cooled bolometers that are used in infrared astronomy. Its characteristics at 300, 77, and 4.2 K are presented. It is also shown to have useful application with certain photoconductive and photovoltaic infrared detectors.

Stability of test environments for performance evaluation of materials for the modular high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Edgemon, G.L.; Wilson, D.F.; Bell, G.E.C.

1993-01-01

Stability of the primary helium-based coolant test gas for use in performance ests of materials for the Modular High-Temperature Gas-Cooled Reactor (MHTGR) was determined. Results of tests of the initial gas chemistry from General Atomics (GA) at elevated temperatures, and the associated results predicted by the SOLGASMIX trademark modelling package are presented. Results indicated that for this gas composition and at flow rates obtainable in the test loop, 466 ± 24C is the highest temperature that can be maintained without significantly altering the specified gas chemistry. Four additional gas chemistries were modelled using SOLGASMIX trademark

Helium-cooled nuclear reactor

International Nuclear Information System (INIS)

Longton, P.B.; Cowen, H.C.

1975-01-01

In helium cooled HTR's there is a by-pass circuit for cleaning purposes in addition to the main cooling circuit. This is to remove such impurities as hydrogen, methane, carbon monoxide and water from the coolant. In this system, part of the coolant successively flows first through an oxidation bed of copper oxide and an absorption bed of silica gel, then through activated charcoal or a molecular sieve. The hydrogen and carbon monoxide impurities are absorbed and the dry gas is returned to the main cooling circuit. To lower the hydrogen/water ratio without increasing the hydrogen fraction in the main cooling circuit, some of the hydrogen fraction converted into water is added to the cooling circuit. This is done, inter alia, by bypassing the water produced in the oxidation bed before it enters the absorption bed. The rest of the by-pass circuit, however, also includes an absorption bed with a molecular sieve. This absorbs the oxidized carbon monoxide fraction. In this way, such side effects as the formation of additional methane, carburization of the materials of the by-pass circuit or loss of graphite are avoided. (DG/RF) [de

Thermodynamic analysis and economical evaluation of two 310-80 K pre-cooling stage configurations for helium refrigeration and liquefaction cycle

Science.gov (United States)

Zhu, Z. G.; Zhuang, M.; Jiang, Q. F.; Y Zhang, Q.; Feng, H. S.

2017-12-01

In 310-80 K pre-cooling stage, the temperature of the HP helium stream reduces to about 80 K where nearly 73% of the enthalpy drop from room temperature to 4.5 K occurs. Apart from the most common liquid nitrogen pre-cooling, another 310-80 K pre-cooling configuration with turbine is employed in some helium cryoplants. In this paper, thermodynamic and economical performance of these two kinds of 310-80 K pre-cooling stage configurations has been studied at different operating conditions taking discharge pressure, isentropic efficiency of turbines and liquefaction rate as independent parameters. The exergy efficiency, total UA of heat exchangers and operating cost of two configurations are computed. This work will provide a reference for choosing 310-80 K pre-cooling stage configuration during design.

Ceramic BOT type blanket with poloidal helium cooling

International Nuclear Information System (INIS)

Cardella, A.; Daenenr, W.; Iseli, M.; Ferrari, M.; Gallina, M.; Rado, V.; Simbolotti, G.; Violante, V.

1989-01-01

This paper briefly describes the work done and results achieved over the past two years on the ceramic breeder BOT blanket with poloidal helium cooling. A conclusive remark on the brick/plate option described previously is followed by short descriptions of the low and high performance pebble bed options elaborated as alternatives for both NET and DEMO. The results show, togethre with those about the poloidal cooling of the First Wall, good prospects for this blanket type provided that the questions connected wiht an extensive use of beryllium find a satisfactor answer. (author). 5 refs.; 7 figs.; 1 tab

Analysis of pressure drop accidents in high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Kameoka, Toshiyuki

1980-01-01

Research and development are carried out on various problems in order to realize a multi-purpose, high temperature gas-cooled experimental reactor by Japan Atomic Energy Research Institute and others. In the experimental reactor in consideration at present, it is planned to flow helium at 1000 deg C and 40 atm. For the purpose, high temperature heat insulation structures are designed and developed, which insulate heat on the internal surfaces of pressure vessels and pipings. Consideration must be given to these internal heat insulation structures about the various characteristics in the working environmental temperature and pressure conditions, the measures for preventing the by-pass flow due to the formation of gaps and the abnormal leak of heat through the natural convection in the heat insulators and others. In this paper, the experimental results on the rapid pressure reduction characteristics of ceramic fiber heat insulation structures are reported. The ceramic fiber heat insulation structures have the features such as the application to uneven surfaces and penetration parts, the prevention of by-pass flow, and very low permeability. The problem is the restoring force after the high temperature compression. The experiment on rapid pressure reduction due to the accidental release of gas and the results are reported. (Kako, I.)

Is cold better ? - exploring the feasibility of liquid-helium-cooled optics

International Nuclear Information System (INIS)

Assoufid, L.; Mills, D.; Macrander, A.; Tajiri, G.

1999-01-01

Both simulations and recent experiments conducted at the Advanced Photon Source showed that the performance of liquid-nitrogen-cooled single-silicon crystal monochromators can degrade in a very rapid nonlinear fashion as the power and for power density is increased. As a further step towards improving the performance of silicon optics, we propose cooling with liquid helium, which dramatically improves the thermal properties of silicon beyond that of liquid nitrogen and brings the performance of single silicon-crystal-based synchrotrons radiation optics up to the ultimate limit. The benefits of liquid helium cooling as well as some of the associated technical challenges will be discussed, and results of thermal and structural finite elements simulations comparing the performance of silicon monochromators cooled with liquid nitrogen and helium will be given

I. Construction of an ultralow temperature laboratory. II. Thermal relaxation in superfluid helium-3

International Nuclear Information System (INIS)

Neuhauser, B.J.

1986-01-01

The first part of this thesis describes the construction of an ultralow temperature laboratory capable of reaching temperatures below 0.002 K. Continuous refrigeration to 0.012 K is provided by a cold plate/dilution refrigerator system. Single-cycle cooling to 0.002 K is accomplished by adiabatic demagnetization of cerous magnesium nitrate (CMN), a paramagnetic salt. Thermometry is done by measuring the resistance of carbon and germanium sensors, the magnetic susceptibility of lanthanum-diluted CMN, and the anisotropy of gamma radiation from a cobalt-60 nuclear orientation thermometer. Systems have been developed to allow precise control of the temperature and pressure of the liquid helium-3 sample. Measurements of thermal relaxation of liquid helium-3 in the ultralow temperature cell following sudden magnetic cooling of the CMN refrigerant are described. Analysis of the transient response of a thermal model of the cell indicates that the ratio of the time constants immediately below and above the superfluid-to-normal transition temperature provides a close estimate of the ratio of the corresponding helium-3 heat capacities, at least in the superfluid A-phase

The gas-cooled high temperature reactor: perspectives, problems and programmes

International Nuclear Information System (INIS)

Beckurts, K.H.; Engelmann, P.; Erb, D.E.

1977-01-01

For nearly 20 years, extensive research and development programs on Helium-cooled high-temperature reactors (HTR) have been carried out in several countries of the world, in particular in Germany and in the United States. This reactor system offers major potential advantages as a source of electricity or of nuclear process heat: it shows high nuclear fuel conversion efficiency, permitting a better utilization of uranium and in particular of thorium resources; it offers a high degree of inherent nuclear safety and thus a good potential for adoption to very strict safety standards; it permits high-efficiency electricity generation using either the indirect steam or the direct Helium cycle; dry air cooling can be employed without major economic penalties; it permits direct use of the nuclear heat for the production of gaseous or liquid secondary fuels from coal and other fossil fuels or - on a more extended time scale - by thermochemical water splitting. As a result of the longstanding efforts, satisfactory solutions have been found for many of the basic problems of this new reactor system, particularly in the field of high-temperature fuels and materials technology. Three small experimental plants - Peach Bottom in USA, Dragon in England, and AVR in Germany - have been operated successfully over extended periods of time. The AVR is still in operation; since 1974 it has performed satisfactorily with an average gas outlet temperature of 950 0 C. Prototype steam-cycle plants of 300 MW(e) are underway at Fort St. Vrain, USA (full-power operation scheduled for 1977), and at Schmehausen, Germany (scheduled for 1979). Major delays have occured in the construction and commissioning of these plants; they are due to various reasons and do not reveal specific problems of the HTR. Commercial market introduction of the steam-cycle electricity generating system has been attempted, but the first approach has not been successfull. Major effects by both government and industry are

Critical Current and Stability of MgB$_2$ Twisted-Pair DC Cable Assembly Cooled by Helium Gas

CERN Document Server

AUTHOR|(CDS)2069632; Ballarino, Amalia; Yang, Yifeng; Young, Edward Andrew; Bailey, Wendell; Beduz, Carlo

2013-01-01

Long length superconducting cables/bus-bars cooled by cryogenic gases such as helium operating over a wider temperature range are a challenging but exciting technical development prospects, with applications ranging from super-grid transmission to future accelerator systems. With limited existing knowledge and previous experiences, the cryogenic stability and quench protection of such cables are crucial research areas because the heat transfer is reduced and temperature gradient increased compared to liquid cryogen cooled cables. V-I measurements on gas-cooled cables over a significant length are an essential step towards a fully cryogenic stabilized cable with adequate quench protection. Prototype twisted-pair cables using high-temperature superconductor and MgB2 tapes have been under development at CERN within the FP7 EuCARD project. Experimental studies have been carried out on a 5-m-long multiple MgB$_2$ cable assembly at different temperatures between 20 and 30 K. The subcables of the assembly showed sim...

Equation of state of fluid helium at high temperatures and densities

Science.gov (United States)

Cai, Lingcang; Chen, Qifeng; Gu, Yunjun; Zhang, Ying; Zhou, Xianming; Jing, Fuqian

2005-03-01

Hugoniot curves and shock temperatures of gas helium with initial temperature 293 K and three initial pressures 0.6, 1.2, and 5.0 MPa were measured up to 15000 K using a two-stage light-gas gun and transient radiation pyrometer. It was found that the calculated Hugoniot EOS of gas helium at the same initial pressure using Saha equation with Debye-Hückel correction was in good agreement with the experimental data. The curve of the calculated shock wave velocity with the particle velocity of gas helium which is shocked from the initial pressure 5 MPa and temperature 293 K, i.e., the D ≈ u relation, D= C 0+λ u ( uionization degree of the shocked gas helium reaches 10-3.

State of the Art Report for a Bearing for VHTR Helium Circulator

International Nuclear Information System (INIS)

Lee, Jae Seon; Song, Kee Nam; Kim, Yong Wan; Lee, Won Jae

2008-10-01

A helium circulator in a VHTR(Very High Temperature gas-cooled Reactor) plays a core role which translates thermal energy at high temperature from a nuclear core to a steam generator. Helium as a operating coolant circulates a primary circuit in high temperature and high pressure state, and controls thermal output of a nuclear core by controlling flow rate. A helium circulator is the only rotating machinery in a VHTR, and its reliability should be guaranteed for reliable operation of a reactor and stable production of hydrogen. Generally a main helium circulator is installed on the top of a steam generator vessel, and helium is circulated only by a main helium circulator in a normal operation state. An auxiliary or shutdown circulator is installed at the bottom of a reactor vessel, and it is an auxiliary circulator for shutting down a reactor in case of refueling or accelerating cooling down in case of fast cooling. Since a rotating shaft of a helium circulator is supported by bearings, bearings are the important machine elements which determines reliability of a helium circulator and a nuclear reactor. Various types of support bearings have been developed and applied for circulator bearings since 1960s, and it is still developing for developing VHTRs. So it is necessary to review and analyze the current technical state of helium circulator support bearings to develop bearings for Koran developing VHTR helium circulator

Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

International Nuclear Information System (INIS)

Wong, C.P.C.; Nygren, R.E.; Baxi, C.B.; Fogarty, P.; Ghoniem, N.; Khater, H.; McCarthy, K.; Merrill, B.; Nelson, B.; Reis, E.E.; Sharafat, S.; Schleicher, R.; Sze, D.K.; Ulrickson, M.; Willms, S.; Youssef, M.; Zinkel, S.

1999-01-01

Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W-5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. System study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kWh. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study

[Fluid dynamics of supercritical helium within internally cooled cabled superconductors

International Nuclear Information System (INIS)

Van Sciver, S.W.

1995-01-01

The Applied Superconductivity Center of the University of Wisconsin-Madison proposes to conduct research on low temperature helium fluid dynamics as it applies to the cooling of internally cooled cabled superconductors (ICCS). Such conductors are used in fusion reactor designs including most of the coils in ITER. The proposed work is primarily experimental involving measurements of transient and steady state pressure drop in a variety of conductor configurations. Both model and prototype conductors for actual magnet designs will be investigated. The primary goal will be to measure and model the friction factor for these complex geometries. In addition, an effort will be made to study transient processes such as heat transfer and fluid expulsion associated with quench conditions

A robust helium-cooled shield/blanket design for ITER

International Nuclear Information System (INIS)

Wong, C.P.C.; Bourque, R.F.; Baxi, C.B.

1993-11-01

General Atomics Fusion and Reactor Groups have completed a helium-cooled, conceptual shield/blanket design for ITER. The configuration selected is a pressurized tubes design embedded in radially oriented plates. This plate can be made from ferritic steel or from V-alloy. Helium leakage to the plasma chamber is eliminated by conservative, redundant design and proper quality control and inspection programs. High helium pressure at 18 MPa is used to reduce pressure drop and enhance heat transfer. This high gas pressure is believed practical when confined in small diameter tubes. Ample industrial experience exists for safe high gas pressure operations. Inboard shield design is highlighted in this study since the allowable void fraction is more limited. Lithium is used as the thermal contacting medium and for tritium breeding, its safety concerns are minimized by a modular, low inventory design that requires no circulation of the liquid metal for the purpose of heat removal. This design is robust, conservative, reliable, and meets all design goals and requirements. It can also be built with present-day technology

Determination of an instability temperature for alloys in the cooling gas of a high temperature reactor

International Nuclear Information System (INIS)

Grimmer, H.; Grman, D.; Krompholz, K.; Zimmermann, U.; Ullrich, G.

1985-05-01

High temperature alloys designed to be used for components in the primary circuit of a helium cooled high temperature nuclear reactor show massive CO production above a certain temperature, called the instability temperature T/sub i/, which increases with increasing partial pressure of CO in the cooling gas. At p/sub CO/ = 15 microbar, T/sub i/ lies between 900 and 950 degrees C for the four alloys under investigation: T/sub i/ is lowest for the iron base alloy Incoloy 800 H and increases for the nickel base alloys in the order Inconel 617, HDA 230 and Nimonic 86. Measurements of T/sub i/ made at 3 different laboratories were compared and shown to agree for p/sub CO/ 25 microbar, compatible with CO production by a reaction of Cr2O3 with carbides. Some measurements of T/sub i/ on HDA 230 and Nimonic 86 were performed in the course of simulated reactor disturbances. They showed that the oxide layer looses its protective properties above T/sub i/. A highlight of the examinations was the detection of eta-carbides (M6C) with unusual properties. M6C is the only type of carbide occuring in HDA 230. An eta-carbide with a lattice constant of 1088.8 pm had developed at the surface of Nimonic 86 during pre-oxidation before the disturbance simulation. Its composition is estimated at Ni3SiMo2C. Eta-carbides containing Si and especially eta-carbides with lattice constants as low as 1088.8 pm have been described only rarely until now. (author)

Numerical benchmark for the deep-burn modular helium-cooled reactor (DB-MHR)

International Nuclear Information System (INIS)

Taiwo, T. A.; Kim, T. K.; Buiron, L.; Varaine, F.

2006-01-01

Numerical benchmark problems for the deep-burn concept based on the prismatic modular helium-cooled reactor design (a Very High Temperature Reactor (VHTR)) are specified for joint analysis by U.S. national laboratories and industry and the French CEA. The results obtained with deterministic and Monte Carlo codes have been inter-compared and used to confirm the underlying feature of the DB-MHR concept (high transuranics consumption). The results are also used to evaluate the impact of differences in code methodologies and nuclear data files on the code predictions for DB-MHR core physics parameters. The code packages of the participating organizations (ANL and CEA) are found to give very similar results. (authors)

Initial assessment of environmental effects on SiC/SiC composites in helium-cooled nuclear systems

Energy Technology Data Exchange (ETDEWEB)

Contescu, Cristian I [ORNL

2013-09-01

This report summarized the information available in the literature on the chemical reactivity of SiC/SiC composites and of their components in contact with the helium coolant used in HTGR, VHTR and GFR designs. In normal operation conditions, ultra-high purity helium will have chemically controlled impurities (water, oxygen, carbon dioxide, carbon monoxide, methane, hydrogen) that will create a slightly oxidizing gas environment. Little is known from direct experiments on the reactivity of third generation (nuclear grade) SiC/SiC composites in contact with low concentrations of water or oxygen in inert gas, at high temperature. However, there is ample information about the oxidation in dry and moist air of SiC/SiC composites at high temperatures. This information is reviewed first in the next chapters. The emphasis is places on the improvement in material oxidation, thermal, and mechanical properties during three stages of development of SiC fibers and at least two stages of development of the fiber/matrix interphase. The chemical stability of SiC/SiC composites in contact with oxygen or steam at temperatures that may develop in off-normal reactor conditions supports the conclusion that most advanced composites (also known as nuclear grade SiC/SiC composites) have the chemical resistance that would allow them maintain mechanical properties at temperatures up to 1200 1300 oC in the extreme conditions of an air or water ingress accident scenario. Further research is needed to assess the long-term stability of advanced SiC/SiC composites in inert gas (helium) in presence of very low concentrations (traces) of water and oxygen at the temperatures of normal operation of helium-cooled reactors. Another aspect that needs to be investigated is the effect of fast neutron irradiation on the oxidation stability of advanced SiC/SiC composites in normal operation conditions.

Helium-flow measurement using ultrasonic technique

International Nuclear Information System (INIS)

Sondericker, J.H.

1983-01-01

While designing cryogenic instrumentation for the Colliding Beam Accelerator (CBA) helium-distribution system it became clear that accurate measurement of mass flow of helium which varied in temperature from room to sub-cooled conditions would be difficult. Conventional venturi flow meters full scale differential pressure signal would decrease by more than an order of magnitude during cooldown causing unacceptable error at operating temperature. At sub-cooled temperatures, helium would be pumped around cooling loops by an efficient, low head pressure circulating compressor. Additional pressure drop meant more pump work was necessary to compress the fluid resulting in a higher outlet temperature. The ideal mass flowmeter for this application was one which did not add pressure drop to the system, functioned over the entire temperature range, has high resolution and delivers accurate mass flow measurement data. Ultrasonic flow measurement techniques used successfully by the process industry, seemed to meet all the necessary requirements. An extensive search for a supplier of such a device found that none of the commercial stock flowmeters were adaptable to cryogenic service so the development of the instrument was undertaken by the CBA Cryogenic Control and Instrumentation Engineering Group at BNL

Ion temperature anisotropy in high power helium neutral beam fuelling experiments in JET

Energy Technology Data Exchange (ETDEWEB)

Maas, A C; Core, W G.F.; Gerstel, U C; Von Hellermann, M G; Koenig, R W.T.; Marcus, F B [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

1994-07-01

During helium beam fuelling experiments in JET, distinctive anisotropic features have been observed in the velocity distribution function describing both fast and thermal alpha particle populations. During the initial fuelling phase the central helium ion temperature observed perpendicular to the magnetic field is higher than the central electron temperature, while the central helium ion temperature observed parallel to the magnetic field is lower than or equal to the central electron temperature. In order to verify temperature measurements of both perpendicular and parallel lines of sight, other independent methods of deducing the ion temperature are investigated: deuterium ion temperature, deuterium density, comparison with neutron rates and profiles (influence of a possible metastable population of helium). 6 refs., 7 figs.

Radiation Protection Practices during the Helium Circulator Maintenance of the 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10

Directory of Open Access Journals (Sweden)

Chengxiang Guo

2016-01-01

Full Text Available Current radiation protection methodology offers abundant experiences on light-water reactors, but very few studies on high temperature gas-cooled reactor (HTR. To fill this gap, a comprehensive investigation was performed to the radiation protection practices in the helium circulator maintenance of the Chinese 10 MW HTR test module (HTR-10 in this paper. The investigation reveals the unique behaviour of HTR-10’s radiation sources in the maintenance as well as its radionuclide species and presents the radiation protection methods that were tailored to these features. Owing to these practices, the radioactivity level was kept low throughout the maintenance and only low-level radioactive waste was generated. The quantitative analysis further demonstrates that the decontamination efficiency was over 89% for surface contamination and over 34% for γ dose rate and the occupational exposure was much lower than both the limits of regulatory and the exposure levels in comparable literature. These results demonstrate the effectiveness of the reported radiation protection practices, which directly provides hands-on experience for the future HTR-PM reactor and adds to the completeness of the radiation protection methodology.

High power density reactors based on direct cooled particle beds

Science.gov (United States)

Powell, J. R.; Horn, F. L.

Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

Survey on Cooled-Vessel Designs in High Temperature Gas-Cooled Reactors

International Nuclear Information System (INIS)

Kim, Min-Hwan; Lee, Won-Jae

2006-01-01

The core outlet temperature of the coolant in the high temperature gas-cooled reactors (HTGR) has been increased to improve the overall efficiency of their electricity generation by using the Brayton cycle or their nuclear hydrogen production by using thermo-chemical processes. The increase of the outlet temperature accompanies an increase of the coolant inlet temperature. A high coolant inlet temperature results in an increase of the reactor pressure vessel (RPV) operation temperature. The conventional steels, proven vessel material in light water reactors, cannot be used as materials for the RPV in the elevated temperatures which necessitate its design to account for the creep effects. Some ferritic or martensitic steels like 2 1/4Cr-1Mo and 9Cr-1Mo-V are very well established creep resistant materials for a temperature range of 400 to 550 C. Although these materials have been used in a chemical plant, there is limited experience with using these materials in nuclear reactors. Even though the 2 1/4Cr-1Mo steel was used to manufacture the RPV for HTR-10 of Japan Atomic Energy Agency(JAEA), a large RPV has not been manufactured by using this material or 9Cr-1Mo-V steel. Due to not only its difficulties in manufacturing but also its high cost, the JAEA determined that they would exclude these materials from the GTHTR design. For the above reasons, KAERI has been considering a cooled-vessel design as an option for the RPV design of a NHDD plant (Nuclear Hydrogen Development and Demonstration). In this study, we surveyed several HTGRs, which adopt the cooled-vessel concept for their RPV design, and discussed their design characteristics. The survey results in design considerations for the NHDD cooled-vessel design

Nonlinear Power-Level Control of the MHTGR Only with the Feedback Loop of Helium Temperature

Directory of Open Access Journals (Sweden)

Zhe Dong

2013-02-01

Full Text Available Power-level control is a crucial technique for the safe, stable and efficient operation of modular high temperature gas-cooled nuclear reactors (MHTGRs, which have strong inherent safety features and high outlet temperatures. The current power-level controllers of the MHTGRs need measurements of both the nuclear power and the helium temperature, which cannot provide satisfactory control performance and can even induce large oscillations when the neutron sensors are in error. In order to improve the fault tolerance of the control system, it is important to develop a power-level control strategy that only requires the helium temperature. The basis for developing this kind of control law is to give a state-observer of the MHTGR a relationship that only needs the measurement of helium temperature. With this in mind, a novel nonlinear state observer which only needs the measurement of helium temperature is proposed. This observer is globally convergent if there is no disturbance, and has the L2 disturbance attenuation performance if the disturbance is nonzero. The separation principle of this observer is also proven, which denotes that this observer can recover the performance of both globally asymptotic stabilizers and L2 disturbance attenuators. Then, a new dynamic output feedback power-level control strategy is established, which is composed of this observer and the well-built static state-feedback power-level control based upon iterative dissipation assignment (IDA-PLC. Finally, numerical simulation results show the high performance and feasibility of this newly-built dynamic output feedback power-level controller.

Development Status of the Helium Circulator for the HCS of HCCR-TBS

International Nuclear Information System (INIS)

Lee, Eo Hwak; Jin, Hyung Gon; Yoon, Jae Sung; Kim, Suk Kwon; Lee, Dong Won; Lee, Si Woo; Cho, Seung Yon

2016-01-01

The calculated eddy current loss on the stainless steel sealing cap of the magnetic coupling device is very high. To solve the eddy current loss problem of the sealing cap, a glass fiber composite, non-conductive and high strength material, is adapted as a material of the sealing cap. The HCCR TBM will be cooled down by HCS (Helium Cooling System), supply high pressure (8 MPa) and temperature (300 .deg. C) helium coolant with 1.15 kg/s of mass flow for nominal operation. The real-scale helium circulator, which is main component of the HCS, has been developed since 2014. In present study, design and manufacture progress of the helium circulator and its verification test plan are described. The real-scale circulator has been developed to provide high temperature and pressure of helium flow as a coolant of the HCCR TBM. To prevent helium leakage, magnetic coupling design was adapted between the shaft and the impeller

Liquid helium cooling of the MFTF superconducting magnets

International Nuclear Information System (INIS)

VanSant, J.H.; Zbasnik, J.P.

1986-09-01

During acceptance testing of the Mirror Fusion Test Facility (MFTF), we measured these tests: liquid helium heat loads and flow rates in selected magnets. We used the data from these tests to estimate helium vapor quality in the magnets so that we could determine if adequate conductor cooling conditions had occurred. We compared the measured quality and flow with estimates from a theoretical model developed for the MFTF magnets. The comparison is reasonably good, considering influences that can greatly affect these values. This paper describes the methods employed in making the measurements and developing the theoretical estimates. It also describes the helium system that maintained the magnets at required operating conditions

Analysis of multi-scale spatial separation in a block-type thorium-loaded helium-cooled high-temperature reactor

International Nuclear Information System (INIS)

Huang, Jie; Ding, Ming

2017-01-01

Highlights: • Four-level of spatial separation is described in a block-type thorium-loaded HTR. • A traditional two-step calculation scheme is used to get the neutronic performance. • Fuel cycle cost is calculated by the levelised lifetime cost method. • Fuel cycle cost decreases with the increase of separation level or thorium content. • Effective enrichment basically determines the fuel cycle cost. - Abstract: With nuclear energy’s rapid development in recent years, supply of nuclear fuel has become increasingly important. Thorium has re-gained attention because of its abundant reserves and excellent physical properties. Compared to the homogeneous Th/U MOX fuel, separation of thorium and uranium in space is a better use of thorium. Therefore, this paper describes four-level spatial separation – no separation, tristructural-isotropic (TRISO) level, channel level and block level – in a block-type thorium-loaded helium-cooled high-temperature reactor (HTR). A traditional two-step calculation scheme, lattice calculation followed by core calculation, is used to get the neutronic performance of the equilibrium cycle, including uranium enrichment, mass of fuel, effective multiplication factor, and average conversion ratio. Based on these data, the fuel cycle cost of different-scale spatial separation can be calculated by the levelised lifetime cost method as a function of thorium content. As the separation level increases from no separation to channel level, the effective enrichment decreases 15% due to the increase of resonance escape probability. So there is a 13% drop for the fuel cycle cost. For TRISO-level separation, as the thorium content increases from 9 to 57%, the effective enrichment decreases 14% because of the superior breeding capacity of U-233. As a result, the fuel cycle cost also has about a 12% decrease. From the perspective of fuel cycle economics, channel-level separation with 60% thorium content is suggested.

Properties of super alloys for high temperature gas cooled reactor

International Nuclear Information System (INIS)

Izaki, Takashi; Nakai, Yasuo; Shimizu, Shigeki; Murakami, Takashi

1975-01-01

The existing data on the properties at high temperature in helium gas of iron base super alloys. Incoloy-800, -802 and -807, nickel base super alloys, Hastelloy-X, Inconel-600, -617 and -625, and a casting alloy HK-40 were collectively evaluated from the viewpoint of the selection of material for HTGRs. These properties include corrosion resistance, strength and toughness, weldability, tube making, formability, radioactivation, etc. Creep strength was specially studied, taking into consideration the data on the creep characteristics in the actual helium gas atmosphere. The necessity of further long run creep data is suggested. Hastelloy-X has completely stable corrosion resistance at high temperature in helium gas. Incoloy 800 and 807 and Inconel 617 are not preferable in view of corrosion resistance. The creep strength of Inconel 617 extraporated to 1,000 deg C for 100,000 hours in air was the greatest rupture strength of 0.6 kg/mm 2 in all above alloys. However, its strength in helium gas began to fall during a relatively short time, so that its creep strength must be re-evaluated in the use for long time. The radioactivation and separation of oxide film in primary construction materials came into question, Inconel 617 and Incoloy 807 showed high induced radioactivity intensity. Generally speaking, in case of nickel base alloys such as Hastelloy-X, oxide film is difficult to break away. (Iwakiri, K.)

Helium leak and chemical impurities control technology in HTTR

International Nuclear Information System (INIS)

Tochio, Daisuke; Shimizu, Atsushi; Hamamoto, Shimpei; Sakaba, Nariaki

2014-01-01

Japan Atomic Energy Agency (JAEA) has designed and developed high-temperature gas-cooled reactor (HTGR) hydrogen cogeneration system named gas turbine high-temperature reactor (GTHTR300C) as a commercial HTGR. Helium gas is used as the primary coolant in HTGR. Helium gas is easy to leak, and the primary helium leakage should be controlled tightly from the viewpoint of preventing the release of radioactive materials to the environment. Moreover from the viewpoint of preventing the oxidization of graphite and metallic material, the helium coolant chemistry should be controlled tightly. The primary helium leakage and the helium coolant chemistry during the operation is the major factor in the HTGR for commercialization of HTGR system. This paper shows the design concept and the obtained operational experience on the primary helium leakage control and primary helium impurity control in the high-temperature engineering test reactor (HTTR) of JAEA. Moreover, the future plan to obtain operational experience of these controls for commercialization of HTGR system is shown. (author)

Helium pressures in RHIC vacuum cryostats and relief valve requirements from magnet cooling line failure

Energy Technology Data Exchange (ETDEWEB)

Liaw, C.J.; Than, Y.; Tuozzolo, J.

2011-03-28

A catastrophic failure of the RHIC magnet cooling lines, similar to the LHC superconducting bus failure incident, would pressurize the insulating vacuum in the magnet and transfer line cryostats. Insufficient relief valves on the cryostats could cause a structural failure. A SINDA/FLUINT{reg_sign} model, which simulated the 4.5K/4 atm helium flowing through the magnet cooling system distribution lines, then through a line break into the vacuum cryostat and discharging via the reliefs into the RHIC tunnel, had been developed to calculate the helium pressure inside the cryostat. Arc flash energy deposition and heat load from the ambient temperature cryostat surfaces were included in the simulations. Three typical areas: the sextant arc, the Triplet/DX/D0 magnets, and the injection area, had been analyzed. Existing relief valve sizes were reviewed to make sure that the maximum stresses, caused by the calculated maximum pressures inside the cryostats, did not exceed the allowable stresses, based on the ASME Code B31.3 and ANSYS results. The conclusions are as follows: (1) The S/F simulation results show that the highest internal pressure in the cryostats, due to the magnet line failure, is {approx}37 psig (255115 Pa); (2) Based on the simulation, the temperature on the cryostat chamber, INJ Q8-Q9, could drop to 228 K, which is lower than the material minimum design temperature allowed by the Code; (3) Based on the ASME Code and ANSYS results, the reliefs on all the cryostats inside the RHIC tunnel are adequate to protect the vacuum chambers when the magnet cooling lines fail; and (4) In addition to the pressure loading, the thermal deformations, due to the temperature decrease on the cryostat chambers, could also cause a high stress on the chamber, if not properly supported.

Helium pressures in RHIC vacuum cryostats and relief valve requirements from magnet cooling line failure

International Nuclear Information System (INIS)

Liaw, C.J.; Than, Y.; Tuozzolo, J.

2011-01-01

A catastrophic failure of the RHIC magnet cooling lines, similar to the LHC superconducting bus failure incident, would pressurize the insulating vacuum in the magnet and transfer line cryostats. Insufficient relief valves on the cryostats could cause a structural failure. A SINDA/FLUINT(reg s ign) model, which simulated the 4.5K/4 atm helium flowing through the magnet cooling system distribution lines, then through a line break into the vacuum cryostat and discharging via the reliefs into the RHIC tunnel, had been developed to calculate the helium pressure inside the cryostat. Arc flash energy deposition and heat load from the ambient temperature cryostat surfaces were included in the simulations. Three typical areas: the sextant arc, the Triplet/DX/D0 magnets, and the injection area, had been analyzed. Existing relief valve sizes were reviewed to make sure that the maximum stresses, caused by the calculated maximum pressures inside the cryostats, did not exceed the allowable stresses, based on the ASME Code B31.3 and ANSYS results. The conclusions are as follows: (1) The S/F simulation results show that the highest internal pressure in the cryostats, due to the magnet line failure, is ∼37 psig (255115 Pa); (2) Based on the simulation, the temperature on the cryostat chamber, INJ Q8-Q9, could drop to 228 K, which is lower than the material minimum design temperature allowed by the Code; (3) Based on the ASME Code and ANSYS results, the reliefs on all the cryostats inside the RHIC tunnel are adequate to protect the vacuum chambers when the magnet cooling lines fail; and (4) In addition to the pressure loading, the thermal deformations, due to the temperature decrease on the cryostat chambers, could also cause a high stress on the chamber, if not properly supported.

Real time thermal hydraulic model for high temperature gas-cooled reactor core

International Nuclear Information System (INIS)

Sui Zhe; Sun Jun; Ma Yuanle; Zhang Ruipeng

2013-01-01

A real-time thermal hydraulic model of the reactor core was described and integrated into the simulation system for the high temperature gas-cooled pebble bed reactor nuclear power plant, which was developed in the vPower platform, a new simulation environment for nuclear and fossil power plants. In the thermal hydraulic model, the helium flow paths were established by the flow network tools in order to obtain the flow rates and pressure distributions. Meanwhile, the heat structures, representing all the solid heat transfer elements in the pebble bed, graphite reflectors and carbon bricks, were connected by the heat transfer network in order to solve the temperature distributions in the reactor core. The flow network and heat transfer network were coupled and calculated in real time. Two steady states (100% and 50% full power) and two transients (inlet temperature step and flow step) were tested that the quantitative comparisons of the steady results with design data and qualitative analysis of the transients showed the good applicability of the present thermal hydraulic model. (authors)

Use of thorium for high temperature gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Guimarães, Cláudio Q., E-mail: claudio_guimaraes@usp.br [Universidade de São Paulo (USP), SP (Brazil). Instituto de Física; Stefani, Giovanni L. de, E-mail: giovanni.stefani@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Santos, Thiago A. dos, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil)

2017-07-01

The HTGR ( High Temperature Gas-cooled Reactor) is a 4{sup th} generation nuclear reactor and is fuelled by a mixture of graphite and fuel-bearing microspheres. There are two competitive designs of this reactor type: The German “pebble bed” mode, which is a system that uses spherical fuel elements, containing a graphite-and-fuel mixture coated in a graphite shell; and the American version, whose fuel is loaded into precisely located graphite hexagonal prisms that interlock to create the core of the vessel. In both variants, the coolant consists of helium pressurised. The HTGR system operates most efficiently with the thorium fuel cycle, however, so relatively little development has been carried out in this country on that cycle for HTGRs. In the Nuclear Engineering Centre of IPEN (Instituto de Pesquisas Energéticas e Nucleares), a study group is being formed linked to thorium reactors, whose proposal is to investigate reactors using thorium for {sup 233}U production and rejects burning. The present work intends to show the use of thorium in HTGRs, their advantages and disadvantages and its feasibility. (author)

Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

International Nuclear Information System (INIS)

Forsberg, C.W.; Conklin, J.; Reich, W.J.

1994-04-01

A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS's heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis

Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

Energy Technology Data Exchange (ETDEWEB)

Forsberg, C.W.; Conklin, J.; Reich, W.J.

1994-04-01

A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

Testing and analyses of a high temperature duct for gas-cooled reactors

International Nuclear Information System (INIS)

Black, W.E.; Roberge, A.; Felten, P.; Bastien, D.

1979-01-01

A 0.6 scale model of a steam cycle gas-cooled reactor high temperature duct was tested in a closed loop helium facility. The object of the test series was to determine: 1) the thermal effects of gas permeation within the thermal barrier, 2) the plastic deformation of the metallic components, and 3) the thermal performance of the fibrous insulation. A series of tests was performed with thermal cyclings from 100 0 C to 760 0 C at 50 atmospheres until the system thermal performance had stabilized hence enabling predictions for the reactor life. Additional tests were made to assess permeation by deliberately simulating sealing weld failures thereby allowing gas flow by-pass within the primary thermal barrier. After 100 cycles the entire primary structure was found to have performed without structural failure. Due to high pressures exerted by the insulation on the cover plates and a design oversight, the thin seal sheets were unable to expand in an anticipated manner. Local buckling resulted. The insulation retained an acceptable degree of resiliency. However, some fiber damage was observed within both the high and low temperature insulation blankets. A thermal analysis was conducted to correlate the hot duct heat transfer results with those obtained from the analytical techniques used for the HTGR design using a computer thermal model representative of the duct and test setup. The thermal performance of the insulation, the temperature gradient through the structural components, the heating load to the cooling system and the permeation flow effect on heat transfer were verified. Exellent correlation between the experimental data and the analytical techniques were obtained

Potential applications of helium-cooled high-temperature reactors to process heat use

International Nuclear Information System (INIS)

Gambill, W.R.; Kasten, P.R.

1981-01-01

High-Temperature Gas-Cooled Reactors (HTRs) permit nuclear energy to be applied to a number of processes presently utilizing fossil fuels. Promising applications of HTRs involve cogeneration, thermal energy transport using molten salt systems, steam reforming of methane for production of chemicals, coal and oil shale liquefaction or gasification, and - in the longer term - energy transport using a chemical heat pipe. Further, HTRs might be used in the more distant future as the energy source for thermochemical hydrogen production from water. Preliminary results of ongoing studies indicate that the potential market for Process Heat HTRs by the year 2020 is about 150 to 250 GW(t) for process heat/cogeneration application, plus approximately 150 to 300 GW(t) for application to fossil conversion processes. HTR cogeneration plants appear attractive in the near term for new industrial plants using large amounts of process heat, possibly for present industrial plants in conjunction with molten-salt energy distribution systems, and also for some fossil conversion processes. HTR reformer systems will take longer to develop, but are applicable to chemicals production, a larger number of fossil conversion processes, and to chemical heat pipes

Research and development for the high-temperature helium-leak detection system (Joint research). Part 2. Development of temperature sensors using optical fibre for the HTTR

Energy Technology Data Exchange (ETDEWEB)

Sakaba, Nariaki; Nakazawa, Toshio; Kawasaki, Kozo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Urakami, Masao; Saisyu, Sadanori [Japan Atomic Power Co., Tokyo (Japan)

2003-03-01

In the second stage of the research and development for a high-temperature helium-leak detection system, the temperature sensor using optical fibres was studied. The sensor detects the helium leakage by the temperature increase surrounded optical fibre with or without heat insulator. Moreover, the applicability of high temperature equipments as the HTTR system was studied. With the sensor we detected 5.0-20.0 cm{sup 3}/s helium leakages within 60 minutes. Also it was possible to detect earlier when the leakage level is at 20.0 cm {sup 3}/s. (author)

Fuel arrangement for high temperature gas cooled reactor

International Nuclear Information System (INIS)

Tobin, J.M.

1978-01-01

Disclosed is a fuel arrangement for a high temperature gas cooled reactor including fuel assemblies with separate directly cooled fissile and fertile fuel elements removably inserted in an elongated moderator block also having a passageway for control elements

Evaluation of Indirect Combined Cycle in Very High Temperature Gas--Cooled Reactor

International Nuclear Information System (INIS)

Chang Oh; Robert Barner; Cliff Davis; Steven Sherman; Paul Pickard

2006-01-01

The U.S. Department of Energy and Idaho National Laboratory are developing a very high temperature reactor to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is twofold: (a) efficient, low-cost energy generation and (b) hydrogen production. Although a next-generation plant could be developed as a single-purpose facility, early designs are expected to be dual purpose, as assumed here. A dual-purpose design with a combined cycle of a Brayton top cycle and a bottom Rankine cycle was investigated. An intermediate heat transport loop for transporting heat to a hydrogen production plant was used. Helium, CO2, and a helium-nitrogen mixture were studied to determine the best working fluid in terms of the cycle efficiency. The relative component sizes were estimated for the different working fluids to provide an indication of the relative capital costs. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the cycle were performed to determine the effects of varying conditions in the cycle. This gives some insight into the sensitivity of the cycle to various operating conditions as well as trade-offs between efficiency and component size. Parametric studies were carried out on reactor outlet temperature, mass flow, pressure, and turbine cooling

Performance of a conduction-cooled high-temperature superconducting bearing

International Nuclear Information System (INIS)

Strasik, M.; Hull, J.R.; Johnson, P.E.; Mittleider, J.; McCrary, K.E.; McIver, C.R.; Day, A.C.

2008-01-01

We report rotational loss measurements for a high-temperature superconducting (HTS) bearing whose cooling consists of a thermal conduction path to the cold head of a cryocooler. Losses have been measured for rotational rates up to 14,500 rpm at different HTS temperatures. The rotational losses decrease with decreasing HTS temperature. For temperatures that can be obtained in a liquid-nitrogen thermosiphon system, at a given speed and gap, the loss of the conduction-cooled HTS bearing is not significantly higher than the loss of a nearly identical HTS bearing cooled by flowing nitrogen from the thermosiphon

Pressurized-helium breakdown at very low temperatures

Energy Technology Data Exchange (ETDEWEB)

Metas, R J

1972-06-01

An investigation of the electrical-breakdown behavior of helium at very low temperatures has been carried out to assist the design and development of superconducting power cables. At very high densities, both liquid and gaseous helium showed an enhancement in electric strength when pressurized to a few atmospheres; conditioned values of breakdown fields then varied between 30 and 45 MV/m. Breakdown processes occurring over a wide range of helium densities are discussed. 24 references.

Dynamics and inherent safety features of small modular high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

1986-01-01

Investigations were made at Oak Ridge National Laboratory to characterize the dynamics and inherent safety features of various modular high temperature gas-cooled reactor (HTGR) designs. This work was sponsored by the US Nuclear Regulatory Commission's HTGR Safety Research program. The US Department of Energy (DOE) and the Gas Cooled Reactor Associates (GCRA) have sponsored studies of several modular HTGR concepts, each having it own unique advantageous economic and inherent safety features. The DOE design team has recently choses a 350-MW(t) annular core with prismatic, graphite matrix fuel for its reference plant. The various safety features of this plant and of the pebble-bed core designs similar to those currently being developed and operated in the Federal Republic of Germany (FRG) are described. A varity of postulated accident sequences involving combinations of loss of forced circulation of the helium primary coolant, loss of primary coolant pressurization, and loss of normal and backup heat sinks were studied and are discussed. Results demonstrate that each concept can withstand an uncontrolled heatup accident without reaching excessive peak fuel temperatures. Comparisons of calculated and measured response for a loss of forced circulation test on the FRG reactor, AVR, are also presented. 10 refs

The high-temperature helium test facility (HHV)

International Nuclear Information System (INIS)

Noack, G.; Weiskopf, H.

1977-03-01

The report describes the high-temperature helium test facility (HHV). Construction of this plant was started in 1972 by Messrs. BBC, Mannheim, on behalf of the Kernforschungsanlage Juelich. By the end of 1976, the construction work is in its last stage, so that the plant may start operation early in 1977. First of all, the cycle system and the arrangement of components are dealt with, followed by a discussion of individual components. Here, emphasis is laid on components typical for HHT systems, while conventional components are mentioned without further structural detail. The projected test programme for the HHV facility in phase IB of the HHT project is shortly dealt with. After this, the potential of this test facility with regard to the possible use of test components and to fluid- and thermodynamic boundary conditions is pointed out. With the unique potential the facility offers here, aspects of shortened service life at higher cycle temperatures do not remain disregarded. (orig./UA) [de

Cryo-cooled high-power window for high-frequency plasma heating

International Nuclear Information System (INIS)

Norajitra, P.; Bojarsky, E.; Reiser, H.; Haefner, H.E.

1991-09-01

Within the framework of gyrotron window development, IMF pursues the concept of a single-disk window with edge cooling. Compared to a double-disk window with surface cooling, this concept offers a number of advantages in terms of safety, reliability, and ease of design and fabrication. The coolants which can be used for this purpose are liquid nitrogen and helium at very low temperature. (orig.) [de

ITER cooling systems

International Nuclear Information System (INIS)

Natalizio, A.; Hollies, R.E.; Sochaski, R.O.; Stubley, P.H.

1992-06-01

The ITER reference system uses low-temperature water for heat removal and high-temperature helium for bake-out. As these systems share common equipment, bake-out cannot be performed until the cooling system is drained and dried, and the reactor cannot be started until the helium has been purged from the cooling system. This study examines the feasibility of using a single high-temperature fluid to perform both heat removal and bake-out. The high temperature required for bake-out would also be in the range for power production. The study examines cost, operational benefits, and impact on reactor safety of two options: a high-pressure water system, and a low-pressure organic system. It was concluded that the cost savings and operational benefits are significant; there are no significant adverse safety impacts from operating either the water system or the organic system; and the capital costs of both systems are comparable

The primary circuit of the dragon high temperature reactor experiment

International Nuclear Information System (INIS)

Simon, R.

2005-01-01

The 20 MWth Dragon Reactor Experiment was the first HTGR (High Temperature Gas-cooled Reactor) with coated particle fuel. Its purpose was to test fuel and materials for the High Temperature Reactor programmes pursued in Europe 40 years ago. This paper describes the design and construction of the primary (helium) circuit. It summarizes the main design objectives, lists the performance data and explains the flow paths of the heat removal and helium purification systems. The principal circuit accidents postulated are discussed and the choice of the main construction materials is given. (author)

Scaling analysis of the coupled heat transfer process in the high-temperature gas-cooled reactor core

International Nuclear Information System (INIS)

Conklin, J.C.

1986-08-01

The differential equations representing the coupled heat transfer from the solid nuclear core components to the helium in the coolant channels are scaled in terms of representative quantities. This scaling process identifies the relative importance of the various terms of the coupled differential equations. The relative importance of these terms is then used to simplify the numerical solution of the coupled heat transfer for two bounding cases of full-power operation and depressurization from full-system operating pressure for the Fort St. Vrain High-Temperature Gas-Cooled Reactor. This analysis rigorously justifies the simplified system of equations used in the nuclear safety analysis effort at Oak Ridge National Laboratory

High temperature gas cooled nuclear reactor

International Nuclear Information System (INIS)

Hosegood, S.B.; Lockett, G.E.

1975-01-01

For high-temperature gas cooled reactors it is considered advantageous to design the core so that the moderator blocks can be removed and replaced by some means of standpipes normally situated in the top of the reactor vessel. An arrangement is here described to facilitate these operations. The blocks have end faces shaped as irregular hexagons with three long sides of equal length and three short sides also of equal length, one short side being located between each pair of adjacent long sides, and the long sides being inclined towards one another at 60 0 . The block defines a number of coolant channels located parallel to its sides. Application of the arrangement to a high temperature gas-cooled reactor with refuelling standpipes is described. The standpipes are located in the top of the reactor vessel above the tops of the columns and are disposed coaxially above the hexagonal channels, with diameters that allow the passage of the blocks. (U.K.)

High-Temperature Structural Analysis Model of the Process Heat Exchanger for Helium Gas Loop (II)

International Nuclear Information System (INIS)

Song, Kee Nam; Lee, Heong Yeon; Kim, Chan Soo; Hong, Seong Duk; Park, Hong Yoon

2010-01-01

PHE (Process Heat Exchanger) is a key component required to transfer heat energy of 950 .deg. C generated in a VHTR (Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute established the helium gas loop for the performance test of components, which are used in the VHTR, and they manufactured a PHE prototype to be tested in the loop. In this study, as part of the high temperature structural-integrity evaluation of the PHE prototype, which is scheduled to be tested in the helium gas loop, we carried out high-temperature structural-analysis modeling, thermal analysis, and thermal expansion analysis of the PHE prototype. The results obtained in this study will be used to design the performance test setup for the PHE prototype

Low-temperature centrifugal helium compressor

International Nuclear Information System (INIS)

Kawada, M.; Togo, S.; Akiyama, Y.; Wada, R.

1974-01-01

A centrifugal helium compressor with gas bearings, which can be operated at the temperature of liquid nitrogen, has been investigated. This compressor has the advantages that the compression ratio should be higher than the room temperature operation and that the contamination of helium could be eliminated. The outer diameter of the rotor is 112 mm. The experimental result for helium gas at low temperature shows a flow rate of 47 g/s and a compression ratio of 1.2 when the inlet pressure was 1 ata and the rotational speed 550 rev/s. The investigation is now focused on obtaining a compression ratio of 1.5. (author)

Pumped helium system for cooling positron and electron traps to 1.2 K

CERN Document Server

Wrubel, J; Kolthammer, W S; Larochelle, P; McConnell, R; Richerme, P; Grzonka, D; Oelert, W; Sefzick, T; Zielinski, M; Borbely, J S; George, M C; Hessels, E A; Storry, C H; Weel, M; Mullers, A; Walz, J; Speck, A

2011-01-01

Extremely precise tests of fundamental particle symmetries should be possible via laser spectroscopy of trapped antihydrogen ((H) over bar) atoms. (H) over bar atoms that can be trapped must have an energy in temperature units that is below 0.5 K-the energy depth of the deepest magnetic traps that can currently be constructed with high currents and superconducting technology. The number of atoms in a Boltzmann distribution with energies lower than this trap depth depends sharply upon the temperature of the thermal distribution. For example, ten times more atoms with energies low enough to be trapped are in a thermal distribution at a temperature of 1.2 K than for a temperature of 4.2 K. To date, (H) over bar atoms have only been produced within traps whose electrode temperature is 4.2 K or higher. A lower temperature apparatus is desirable if usable numbers of atoms that can be trapped are to eventually be produced. This report is about the pumped helium apparatus that cooled the trap electrodes of an (H) ove...

Effect of carburizing helium environment on creep behavior of Ni-base heat-resistant alloys for high-temperature gas-cooled reactors

International Nuclear Information System (INIS)

Kurata, Yuji; Ogawa, Yutaka; Nakajima, Hajime

1988-01-01

Creep tests were conducted on Ni-base heat-resistant alloys Hastelloy XR and XR-II, i.e. versions of Hastelloy X modified for nuclear applications, at 950degC using four types of helium environment with different impurity compositions, and mainly the effect of carburization was examined. For all the materials tested, the values of creep rupture time obtained under the carburizing conditions were similar to or longer than those in the commonly used, standard test environment (JAERI Type B helium). The difference among the results was interpreted by the counterbalancing effects of the strengthening due to carburization and possible weakening caused under very low oxidizing potential. In the corrosion monitoring specimens pronounced carbon pick-up was observed in the environment with high carbon activity and very low oxidizing potential. Based on the results obtained in the present and the previous works, it is suggested that a moderate control of the impurity chemistry is important rather than simple purification of the coolant in protecting the material from the environment-enhanced degradation. Either condition with high or low extremes in the oxidizing and carburizing potentials may cause enhanced degradation and thus are desirable to be avoided at the elevated temperatures. (author)

Friction and wear studies of graphite and a carbon-carbon composite in air and in helium

International Nuclear Information System (INIS)

Li, C.C.; Sheehan, J.E.

1980-10-01

Sliding friction and wear tests were conducted on a commercial isotropic graphite and a carbon-carbon composite in air, purified helium, and a helium environment containing controlled amounts of impurities simulating the primary coolant chemistry of a high-temperature gas-cooled reactor (HTGR). The friction and wear characteristics of the materials investigated were stable and were found to be very sensitive to the testing temperature. In general, friction and wear decreased with increasing temperature in the range from ambient to 950 0 C. This temperature dependence is concluded to be due to chemisorption of impurities to form lubricating films and oxidation at higher temperatures, which reduce friction and wear. Graphite and carbon-carbon composites are concluded to be favorable candidate materials for high-temperature sliding service in helium-cooled reactors

Operating Manual of Helium Refrigerator (Rev. 2)

Energy Technology Data Exchange (ETDEWEB)

Song, K.M.; Son, S.H.; Kim, K.S.; Lee, S.K.; Kim, M.S. [Korea Electric Power Research Institute, Taejon (Korea)

2002-07-01

A helium refrigerator was installed as a supplier of 20K cold helium to the cryogenic distillation system of WTRF pilot plant. The operating procedures of the helium refrigerator, helium compressor and auxiliary apparatus are described for the safety and efficient operation in this manual. The function of the helium refrigerator is to remove the impurities from the compressed helium of about 250psig, to cool down the helium from ambient temperature to 20K through the heat exchanger and expansion engine and to transfer the cold helium to the cryogenic distillation system. For the smoothly operation of helium refrigerator, the preparation, the start-up, the cool-down and the shut-down of the helium refrigerator are described in this operating manual. (author). 3 refs., 14 tabs.

Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

Energy Technology Data Exchange (ETDEWEB)

Peterson, Per [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering; Greenspan, Ehud [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering

2015-02-09

This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m³ to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m³. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X

Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

International Nuclear Information System (INIS)

Peterson, Per; Greenspan, Ehud

2015-01-01

This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3 . This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel

Stability analysis of high temperature superconducting coil in liquid hydrogen

International Nuclear Information System (INIS)

Nakayama, T.; Yagai, T.; Tsuda, M.; Hamajima, T.

2007-01-01

Recently, it is expected that hydrogen plays an important role in energy source including electric power in near future. Liquid hydrogen has high potential for cooling down superconducting coil wound with high temperature superconductors (HTS), such as BSCCO, YBCO. In this paper, we study stabilities of the coils wound with BSCCO tapes, which are immersed in the liquid hydrogen, and compare stability results with those cooled by liquid helium. We treat a minimum propagation zone (MPZ) theory to evaluate the coil stability considering boiling heat flux of the liquid hydrogen, and specific heat, heat conduction and resistivity of HTS materials as a function of temperature. It is found that the coil cooled by the liquid hydrogen has higher stability margin than that cooled by the liquid helium. We compare the stability margins of both coils wound with Bi-2223/Ag tape and Bi-2212/Ag tape in liquid hydrogen. As a result, it is found that the stability of Bi-2212 coil is equivalent to that of Bi-2223 coil in low and high magnetic field, while the maximum current of Bi-2212 coil exceeds a little bit that of Bi-2223 coil in both magnetic fields

Study on a method for loading a Li compound to produce tritium using high-temperature gas-cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Nakaya, Hiroyuki, E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, Hideaki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Katayama, Kazunari [Department of Advanced Energy Engineering Science, Kyushu University, 6-1 Kasuga-koen, Kasuga 8168580 (Japan); Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan)

2015-10-15

Highlights: • Tritium production by a high-temperature gas-cooled reactor was studied. • The loading method considering tritium outflow suppression was estimated. • A reactor with 600 MWt produced 400–600 g of tritium for 180 days. • A possibility that tritium outflow can be sufficiently suppressed was shown. - Abstract: Tritium production using high-temperature gas-cooled reactors and its outflow from the region loading Li compound into the helium coolant are estimated when considering the suppression of tritium outflow. A Li rod containing a cylindrical Li compound placed in an Al{sub 2}O{sub 3} cladding tube is assumed as a method for loading Li compound. A gas turbine high-temperature reactor of 300 MW electrical nominal capacity (GTHTR300) with 600 MW thermal output power is considered and modeled using the continuous-energy Monte Carlo transport code MVP-BURN, where burn-up simulations are carried out. Tritium outflow is estimated from equilibrium solution for the tritium diffusion equation in the cladding tube. A GTHTR300 can produce 400–600 g of tritium over a 180-day operation using the chosen method of loading the Li compound while minimizing tritium outflow from the cladding tube. Optimizing tritium production while suppressing tritium outflow is discussed.

Re-Condensation and Liquefaction of Helium and Hydrogen Using Coolers

International Nuclear Information System (INIS)

Green, Michael A.

2009-01-01

Coolers are used to cool cryogen free devices at temperatures from 5 to 30 K. Cryogen free cooling involves a temperature drop within the device being cooled and between the device and the cooler cold heads. Liquid cooling with a liquid cryogen distributed over the surface of a device combined with re-condensation can result in a much lower temperature drop between the cooler and the device being cooled. The next logical step beyond simple re-condensation is using a cooler to liquefy the liquid cryogen in the device. A number of tests of helium liquefaction and re-condensation of helium have been run using a pulse tube cooler in the drop-in mode. This report discusses the parameter space over which re-condensation and liquefaction for helium and hydrogen can occur.

Particle exhaust of helium plasmas with actively cooled outboard pump limiter on Tore Supra

International Nuclear Information System (INIS)

Uckan, T.; Mioduszewski, P.K.; Loarer, T.; Chatelier, M.; Guilhem, D.; Lutz, T.; Nygren, R.E.; Mahdavi, M.A.

1995-08-01

The superconducting tokamak Tore Supra was designed for long-pulse (30-s) high input power operation. Here observations on the particle-handling characteristics of the actively cooled modular outboard pump limiter (OPL) are presented for helium discharges. The important experimental result was that a modest pumping speed (1 m 3 /s) of the OPL turbomolecular pump (TMP) provided background helium exhaust. This result came about due to a well-conditioned vessel wall with helium discharges that caused no wall outgasing. The particle accountability in these helium discharges was excellent, and the well-conditioned wall did not play a significant role in the particle balance. The helium density control, 25% density drop with OPL exhaust efficiency of ∼1%, was possible with TMP although this may not be the case with reactive gases such as deuterium. The observed quadratic increase of the OPL neutral pressure with helium density was consistent with an improvement of the particle control with increasing plasma density

Hydrogen production system based on high temperature gas cooled reactor energy using the sulfur-iodine (SI) thermochemical water splitting cycle

International Nuclear Information System (INIS)

Garcia, L.; Gonzalez, D.

2011-01-01

Hydrogen production from water using nuclear energy offers one of the most attractive zero-emission energy strategies and the only one that is practical on a substantial scale. Recently, strong interest is seen in hydrogen production using heat of a high-temperature gas-cooled reactor. The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using thermochemical or high-temperature electrolysis (HTE) processes. Eventually it could be also employ a high-temperature gas-cooled reactor (HTGR), which is particularly attractive because it has unique capability, among potential future generation nuclear power options, to produce high-temperature heat ideally suited for nuclear-heated hydrogen production. Using heat from nuclear reactors to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been interest of many laboratories in the world. One of the promising approaches to produce large quantity of hydrogen in an efficient way using the nuclear energy is the sulfur-iodine (SI) thermochemical water splitting cycle. Among the thermochemical cycles, the sulfur iodine process remains a very promising solution in matter of efficiency and cost. This work provides a pre-conceptual design description of a SI-Based H2-Nuclear Reactor plant. Software based on chemical process simulation (CPS) was used to simulate the thermochemical water splitting cycle Sulfur-Iodine for hydrogen production. (Author)

NASA space applications of high-temperature superconductors

Science.gov (United States)

Heinen, Vernon O.; Sokoloski, Martin M.; Aron, Paul R.; Bhasin, Kul B.

1992-01-01

The application of superconducting technology in space has been limited by the requirement of cooling to near liquid helium temperatures. The only means of attaining these temperatures has been with cryogenic fluids which severely limits mission lifetime. The development of materials with superconducting transition temperatures (T sub c) above 77 K has made superconducting technology more attractive and feasible for employment in aerospace systems. Potential applications of high-temperature superconducting technology in cryocoolers and remote sensing, communications, and power systems are discussed.

Updated conceptual design of helium cooling ceramic blanket for HCCB-DEMO

Energy Technology Data Exchange (ETDEWEB)

Wang, Suhao [University of Science and Technology of China, Hefei, Anhui (China); Southwestern Institute of Physics, Chengdu, Sichuan (China); Cao, Qixiang; Wu, Xinghua; Wang, Xiaoyu; Zhang, Guoshu [Southwestern Institute of Physics, Chengdu, Sichuan (China); Feng, Kaiming, E-mail: fengkm@swip.ac.cn [Southwestern Institute of Physics, Chengdu, Sichuan (China)

2016-11-15

Highlights: • An updated design of Helium Cooled Ceramic breeder Blanket (HCCB) for HCCB-DEMO is proposed in this paper. • The Breeder Unit is transformed to TBM-like sub-modules, with double “banana” shape tritium breeder. Each sub-module is inserted in space formed by Stiffen Grids (SGs). • The performance analysis is performed based on the R&D development of material, fabrication technology and safety assessment in CN ITER TBM program. • Hot spots will be located at the FW bend side. - Abstract: The basic definition of the HCCB-DEMO plant and preliminary blanket designed by Southwestern Institution of Physics was proposed in 2009. The DEMO fusion power is 2550 MW and electric power is 800 MW. Based on development of R&D in breeding blanket, a conceptual design of helium cooled blanket with ceramic breeder in HCCB-DEMO was presented. The main design features of the HCCB-DEMO blanket were: (1) CLF-1 structure materials, Be multiplier and Li{sub 4}SiO{sub 4} breeder; (2) neutronic wall load is 2.3 MW/m{sup 2} and surface heat flux is 0.43 MW/m{sup 2} (2) TBR ≈ 1.15; (3) geometry of breeding units is ITER TBM-like segmentation; (4)Pressure of helium is 8 MPa and inlet/outlet temperature is 300/500 °C. On the basis of these design, some important analytical results are presented in aspects of (i) neutronic behavior of the blanket; (ii) design of 3D structure and thermal-hydraulic lay-out for breeding blanket module; (iii) structural-mechanical behavior of the blanket under pressurization. All of these assessments proved current stucture fulfill the design requirements.

Helium turbomachinery operating experience from gas turbine power plants and test facilities

International Nuclear Information System (INIS)

McDonald, Colin F.

2012-01-01

The closed-cycle gas turbine, pioneered and deployed in Europe, is not well known in the USA. Since nuclear power plant studies currently being conducted in several countries involve the coupling of a high temperature gas-cooled nuclear reactor with a helium closed-cycle gas turbine power conversion system, the experience gained from operated helium turbomachinery is the focus of this paper. A study done as early as 1945 foresaw the use of a helium closed-cycle gas turbine coupled with a high temperature gas-cooled nuclear reactor, and some two decades later this was investigated but not implemented because of lack of technology readiness. However, the first practical use of helium as a gas turbine working fluid was recognized for cryogenic processes, and the first two small fossil-fired helium gas turbines to operate were in the USA for air liquefaction and nitrogen production facilities. In the 1970's a larger helium gas turbine plant and helium test facilities were built and operated in Germany to establish technology bases for a projected future high efficiency large nuclear gas turbine power plant concept. This review paper covers the experience gained, and the lessons learned from the operation of helium gas turbine plants and related test facilities, and puts these into perspective since over three decades have passed since they were deployed. An understanding of the many unexpected events encountered, and how the problems, some of them serious, were resolved is important to avoid them being replicated in future helium turbomachines. The valuable lessons learned in the past, in many cases the hard way, particularly from the operation in Germany of the Oberhausen II 50 MWe helium gas turbine plant, and the technical know-how gained from the formidable HHV helium turbine test facility, are viewed as being germane in the context of current helium turbomachine design work being done for future high efficiency nuclear gas turbine plant concepts. - Highlights: �

Conceptual design and analysis of the helium cooled solid breeder blanket for CFETR

Energy Technology Data Exchange (ETDEWEB)

Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Li, Min; Lv, Zhongliang; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Wang, Xiaoliang; Zheng, Jie; Ye, Minyou

2015-10-15

Highlights: • A helium cooled solid blanket was proposed as a candidate blanket concept for CFETR. • Material selection, basic structure and gas flow scheme of the blanket were introduced. • A series of performance analyses for the blanket were summarized. - Abstract: To bridge the gap between ITER and DEMO and to realize the fusion energy in China, a fusion device Chinese Fusion Engineering Test Reactor (CFETR) was proposed and is being designed mainly to demonstrate 50–200 MW fusion power, 30–50% duty time factor, tritium self-sustained. Because of the high demand of tritium production and the realistic engineering consideration, the design of tritium breeding blanket for CFETR is a challenging work and getting special attention. As a blanket candidate, a helium cooled solid breeder blanket has been designed with the emphasis on conservative design and realistic blanket technology. This paper introduces the basic blanket scheme, including the material selection, structural design, cooling scheme and purge gas flow path. In addition, some results of neutronics, thermal-hydraulic and stress analysis are presented.

Optimal thermal-hydraulic performance for helium-cooled divertors

International Nuclear Information System (INIS)

Izenson, M.G.; Martin, J.L.

1996-01-01

Normal flow heat exchanger (NFHX) technology offers the potential for cooling divertor panels with reduced pressure drops (<0.5% Δp/p), reduced pumping power (<0.75% pumping/thermal power), and smaller duct sizes than conventional helium heat exchangers. Furthermore, the NFHX can easily be fabricated in the large sizes required for divertors in large tokamaks. Recent experimental and computational results from a program to develop NFHX technology for divertor coolings using porous metal heat transfer media are described. We have tested the thermal and flow characteristics of porous metals and identified the optimal heat transfer material for the divertor heat exchanger. Methods have been developed to create highly conductive thermal bonds between the porous material and a solid substrate. Computational fluid dynamics calculations of flow and heat transfer in the porous metal layer have shown the capability of high thermal effectiveness. An 18-kW NFHX, designed to meet specifications for the international Thermonuclear Experimental Reactor divertor, has been fabricated and tested for thermal and flow performance. Preliminary results confirm design and fabrication methods. 11 refs., 12 figs., 1 tab

The maximum temperature of a thermodynamic cycle effect on weight-dimensional characteristics of the NPP energy blocks with air cooling

International Nuclear Information System (INIS)

Bezborodov, Yu.A.; Bubnov, V.P.; Nesterenko, V.B.

1982-01-01

The cycle maximum temperature effect on the properties of individual apparatuses and total NPP energy blocks characteristics has been investigated. Air, nitrogen, helium and chemically reacting system N 2 O 4 +2NO+O 2 have been considered as coolants. The conducted investigations have shown that maximum temperature of thermodynamical cycle affects considerably both the weight-dimensional characteristics of individual elements of NPP and total characteristics of NPP energy block. Energy blocks of NPP with air cooling wherein dissociating nitrogen tetroxide is used as working body, have better indexes on the majority of characteristics in comparison with blocks with air, nitrogen and helium cooling. If technical restrictions are to be taken into account (thermal resistance of metals, coolant decomposition under high temperatures, etc.) then dissociating nitrogen tetroxide should be recommended as working body and maximum cycle temperature in the range from 500 up to 600 deg C

RCCS Experiments and Validation for High Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Chang Oh; Cliff Davis; Goon C. Park

2007-01-01

A reactor cavity cooling system (RCCS), an air-cooled helical coil RCCS unit immersed in the water pool, was proposed to overcome the disadvantages of the weak cooling ability of air-cooled RCCS and the complex structure of water-cooled RCCS for the high temperature gas-cooled reactor (HTGR). An experimental apparatus was constructed to investigate the various heat transfer phenomena in the water pool type RCCS, such as the natural convection of air inside the cavity, radiation in the cavity, the natural convection of water in the water pool and the forced convection of air in the cooling pipe. The RCCS experimental results were compared with published correlations. The CFX code was validated using data from the air-cooled portion of the RCCS. The RELAP5 code was validated using measured temperatures from the reactor vessel and cavity walls

Modeling and Simulation of the Sulfur-Iodine Process Coupled to a Very High-Temperature Gas-Cooled Nuclear Reactor

International Nuclear Information System (INIS)

Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan

2015-01-01

Hydrogen produced from water using nuclear energy will avoid both the use of fossil fuel and CO 2 emission presumed to be the dominant reason for global warming. A thermo-chemical sulfur-iodine (SI) process coupled to a Very High Temperature Gas-Cooled Reactor(VHTR) is one of the most prospective hydrogen production methods that split water using nuclear energy because the SI process is suitable for large-scale hydrogen production without CO 2 emission. The dynamic simulation code to evaluate the start-up behavior of the chemical reactors placed on the secondary helium loop of the SI process has been developed and partially verified using the steady state values obtained from the Aspen Plus TM Code simulation. As the start-up dynamic simulation results of the SI process coupled to the IHX, which is one of components in the VHTR system, it is expected that the integrated secondary helium loop of the SI process can be successfully and safely approach the steady state condition

Buffer-gas cooling of antiprotonic helium to 1.5 to 1.7 K, and antiproton-to–electron mass ratio

CERN Document Server

Hori, Masaki; Sótér, Anna; Barna, Daniel; Dax, Andreas; Hayano, Ryugo; Kobayashi, Takumi; Murakami, Yohei; Todoroki, Koichi; Yamada, Hiroyuki; Horváth, Dezső; Venturelli, Luca

2016-01-01

Charge, parity, and time reversal (CPT) symmetry implies that a particle and its antiparticle have the same mass. The antiproton-to-electron mass ratio Embedded Image can be precisely determined from the single-photon transition frequencies of antiprotonic helium. We measured 13 such frequencies with laser spectroscopy to a fractional precision of 2.5 × 10−9 to 16 × 10−9. About 2 × 109 antiprotonic helium atoms were cooled to temperatures between 1.5 and 1.7 kelvin by using buffer-gas cooling in cryogenic low-pressure helium gas; the narrow thermal distribution led to the observation of sharp spectral lines of small thermal Doppler width. The deviation between the experimental frequencies and the results of three-body quantum electrodynamics calculations was reduced by a factor of 1.4 to 10 compared with previous single-photon experiments. From this, Embedded Image was determined as 1836.1526734(15), which agrees with a recent proton-to-electron experimental value within 8 × 10−10.

CARR-CNS with crescent-shape moderator cell and sub-cooling helium jacket surrounding cell

International Nuclear Information System (INIS)

Yu, Qingfeng; Feng, Quanke; Kawai, Takeshi; Shen, Feng; Yuan, Luzheng

2005-01-01

The new type of the moderator cell was developed for the Cold Neutron Source (CNS) of the China Advanced Research Reactor (CARR) which is now constructing at the China Institute of Atomic Energy in Beijing. A crescent-shape moderator cell covered by the sub-cooling helium jacket is adopted. A crescent-shape would help to increase the volume of the moderator cell for corresponding it to the 4 cold neutron guide tubes, even if liquid hydrogen not liquid deuterium were used as a cold moderator. The sub-cooling helium jacket covering the moderator cell removes the nuclear heating of the outer shell wall of the cell. It contributes to reduce the void fraction of liquid hydrogen in the inner shell. Such a type of a moderator cell is suitable for the CNS with higher nuclear heating. The cold helium gas flows down firstly into the sub-cooling helium jacket and then flows up to the condenser. Therefore, the theory of the self-regulation for the thermo-siphon type of the CNS is also applicable

CARR-CNS with crescent-shape moderator cell and sub-cooling helium jacket around cell

International Nuclear Information System (INIS)

Yu, Qingfeng; Feng, Quanke; Kawai, Takeshi; Cheng, Liang; Shen, Feng; Yuan, Luzheng

2005-01-01

The new type of the moderator cell was developed for the Cold Neutron Source (CNS) of the China Advanced Research Reactor (CARR) which is now constructing at the China Institute of Atomic Energy in Beijing. A crescent-shape moderator cell covered by the sub-cooling helium jacket is adopted. A crescent-shape would help to increase the volume of the moderator cell for corresponding it to the 4 cold neutron guide tubes, even if liquid hydrogen not liquid deuterium were used as a cold moderator. The sub-cooling helium jacket covering the moderator cell removes the nuclear heating of the outer shell wall of the cell. It contributes to reduce the void fraction of liquid hydrogen in the inner shell. Such a type of a moderator cell is suitable for the CNS with higher nuclear heating. The cold helium gas flows down firstly into the sub-cooling helium jacket and then flows up to the condenser. Therefore, the theory of the self-regulation for the thermo-siphon type of the CNS is also applicable

Preliminary design of the cooling system for a gas-cooled, high-fluence fast pulsed reactor (HFFPR)

International Nuclear Information System (INIS)

Monteith, H.C.

1978-10-01

The High-Fluence Fast Pulsed Reactor (HFFPR) is a research reactor concept currently being evaluated as a source for weapon effects experimentation and advanced reactor safety experiments. One of the designs under consideration is a gas-cooled design for testing large-scale weapon hardware or large bundles of full-length, fast reactor fuel pins. This report describes a conceptual cooling system design for such a reactor. The primary coolant would be helium and the secondary coolant would be water. The size of the helium-to-water heat exchanger and the water-to-water heat exchanger will be on the order of 0.9 metre (3 feet) in diameter and 3 metres (10 feet) in length. Analysis indicates that the entire cooling system will easily fit into the existing Sandia Engineering Reactor Facility (SERF) building. The alloy Incoloy 800H appears to be the best candidate for the tube material in the helium-to-water heat exchanger. Type 316 stainless steel has been recommended for the shell of this heat exchanger. Estimates place the cost of the helium-to-water heat exchanger at approximately $100,000, the water-to-water heat exchanger at approximately $25,000, and the helium pump at approximately $450,000. The overall cost of the cooling system will approach $2 million

Adsorption pump for helium pumping out

International Nuclear Information System (INIS)

Donde, A.L.; Semenenko, Yu.E.

1981-01-01

Adsorption pump with adsorbent cooling by liquid helium is described. Shuttered shield protecting adsorbent against radiation is cooled with evaporating helium passing along the coil positioned on the shield. The pump is also equipped with primed cylindrical shield, cooled with liquid nitrogen. The nitrogen shield has in the lower part the shuttered shield, on the pump casing there is a valve used for pump pre-burning, and valves for connection to recipient as well. Pumping- out rates are presented at different pressures and temperatures of adsorbent. The pumping-out rate according to air at absorbent cooling with liquid nitrogen constituted 5x10 -4 Pa-3000 l/s, at 2x10 -2 Pa-630 l/s. During the absorbent cooling with liquid hydrogen the pumping-out rate according to air was at 4x10 -4 Pa-580 l/s, at 2x10 -3 Pa-680 l/s, according to hydrogen - at 8x10 -5 Pa-2500 l/s, at 5x10 -3 Pa-4200 l/s. During adsorbent cooling with liquid helium the rate of pumping-out according to hydrogen at 3x10 5 Pa-2400% l/s, at 6x10 3 Pa-1200 l/s, and according to helium at 3.5x10 -5 Pa-2800 l/s, at 4x10 -3 Pa-1150 l/s. The limit vacuum is equal to 1x10 -7 Pa. The volume of the vessel with liquid helium is equal to 3.5 l. Helium consumption is 80 cm 3 /h. Consumption of liquid nitrogen from the shield is 400 cm 3 /h. The limit pressure in the pump is obtained after forevacuum pumping-out (adsorbent regeneration) at 300 K temperature. The pump is made of copper. The pump height together with primed tubes is 800 mm diameter-380 mm [ru

Thermohydraulics design and thermomechanics analysis of two European breeder blanket concepts for DEMO. Pt. 1 and Pt. 2. Pt. 1: BOT helium cooled solid breeding blanket. Pt. 2: Dual coolant self-cooled liquid metal blanket

International Nuclear Information System (INIS)

Norajitra, P.

1995-06-01

Two different breeding blanket concepts are being elaborated at Forschungszentrum Karlsruhe within the framework of the DEMO breeding blanket development, the concept of a helium cooled solid breeding blanket and the concept of a self-cooled liquid metal blanket. The breeder material used in the first concept is Li 4 SiO 4 as a pebble bed arranged separate from the beryllium pebble bed, which serves as multiplier. The breeder material zone is cooled by several toroidally-radially configurated helium cooling plates which, at the same time, act as reinforcements of the blanket structures. In the liquid metal blanket concept lead-lithium is used both as the breeder material and the coolant. It flows at low velocity in poloidal direction downwards and back in the blanket front zone. In both concepts the First Wall is cooled by helium gas. This report deals with the thermohydraulics design and thermomechanics analysis of the two blanket concepts. The performance data derived from the Monte-Carlo computations serve as a basis for the design calculations. The coolant inlet and outlet temperatures are chosen with the design criteria and the economics aspects taken into account. Uniform temperature distribution in the blanket structures can be achieved by suitable branching and routing of the coolant flows which contributes to reducing decisively the thermal stress. The computations were made using the ABAQUS computer code. The results obtained of the stresses have been evaluated using the ASME code. It can be demonstrated that all maximum values of temperature and stress are below the admissible limit. (orig.) [de

The gas-cooled high temperature reactor. Perspectives, problems and programmes

International Nuclear Information System (INIS)

Beckurts, K.H.; Engelmann, P.; Erb, D.E.

1977-01-01

For nearly 20 years extensive research and development programmes on helium-cooled high temperature reactors (HTR) have been carried out in several countries of the world. As a result of the long-standing efforts, satisfactory solutions have been found for many of the basic problems of this new reactor system, particularly in the field of high temperature fuels and materials technology. Three small experimental plants have been operated successfully over extended periods of time. Prototype steam-cycle plants of 300MW(e) are under way at Fort St. Vrain (full-power operation scheduled for 1977) and at Schmehausen (scheduled for 1979). Major delays have occurred in the construction and commissioning of these plants for various reasons but do not reveal specific problems of the HTR. Commercial market introduction of the steam-cycle electricity generating system has been attempted, but the first approach has not been successful. Major efforts both by governments and industry are now required to ensure a successful second approach. To reach competitivity with established nuclear power systems and to take full advantage of the fuel conservation potential of the HTR requires the implementation of the closed thorium fuel cycle on a commercial scale. While some key steps of this cycle have been implemented on a laboratory scale, progress towards a prototype recycling facility has been slow. Closing the thorium fuel cycle represents a major challenge and can only be achieved in a close international collaboration. The paper discusses the world-wide status and potential of HTR technology and reviews the major international development programmes. (author)

IAEA high temperature gas-cooled reactor activities

International Nuclear Information System (INIS)

Kendall, J.M.

2000-01-01

The IAEA activities on high temperature gas-cooled reactors are conducted with the review and support of the Member states, primarily through the International Working Group on Gas-Cooled Reactors (IWG-GCR). This paper summarises the results of the IAEA gas-cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (authors)

IAEA high temperature gas cooled reactor activities

International Nuclear Information System (INIS)

Kendall, J.M.

2001-01-01

IAEA activities on high temperature gas cooled reactors are conducted with the review and support of Member States, primarily through the International Working Group on Gas Cooled Reactors (IWGGCR). This paper summarises the results of the IAEA gas cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products, and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (author)

Gas erosion of impeller housing in the operation of a high-temperature, high-pressure helium circulator

International Nuclear Information System (INIS)

Sanders, J.P.; Heestand, R.L.; Young, H.C.

1988-01-01

Three gas-bearing circulators are installed in series in a high-pressure, high-temperature loop to provide helium flow up to 0.47 m 3 /s at a total head of 78 kJ/kg. The design pressure is 10.7 MPa, and temperatures of 1000 deg. C can be obtained in the test section. The inlet temperature to the circulators is limited to 450 deg. C. The 200-kW motor for each circulator is enclosed in the pressure boundary, and the motor is cooled by circulating the gas within the cavity over a water-cooled coil. The full operating speed is 23,500 rpm. A full-flow filter, absolute for particulate above 10 μm, is installed upstream of the circulator to protect the gas bearing surfaces. The minimum clearances between these surfaces during operation are in the range of 15 to 30 μm. During a routine examination of the circulator, deep V-shaped grooves were found in the stationary surface of this cavity. At the same time, a very fine, dark particulate was observed in crevices of the housing. At first it was assumed that the grooves were formed by particulate erosion; however, examination of the grooves and discussions with persons experienced with large circulator operation changed this opinion. Erosion caused by particulate is characteristically rounded on the bottom and has a greater width to depth aspect than the V-shaped grooves, which were observed. Analysis of the particulate indicated that it was essentially the material of the housing that had undergone reactions with impurities in the circulating gas. It was subsequently concluded that the impeller housing had not been heat treated in a sufficiently oxidizing atmosphere after machining to form an adherent oxide coating. This suboxide coating was eroded by the shear forces in the gas. The exposed layer of metal was then further oxidized by the impurities in the gas, and these layers of oxide were successively eroded to produce the grooves. This erosion problem was eliminated by machining a ring of the same material, heat

Helium temperature measurements in a hot filament magnetic mirror plasma using high resolution Doppler spectroscopy

Science.gov (United States)

Knott, S.; McCarthy, P. J.; Ruth, A. A.

2016-09-01

Langmuir probe and spectroscopic diagnostics are used to routinely measure electron temperature and density over a wide operating range in a reconfigured Double Plasma device at University College Cork, Ireland. The helium plasma, generated through thermionic emission from a negatively biased tungsten filament, is confined by an axisymmetric magnetic mirror configuration using two stacks of NdFeB permanent magnets, each of length 20 cm and diameter 3 cm placed just outside the 15 mm water cooling jacket enclosing a cylindrical vacuum vessel of internal diameter 25 cm. Plasma light is analysed using a Fourier Transform-type Bruker spectrometer with a highest achievable resolution of 0.08 cm-1 . In the present work, the conventional assumption of room temperature ions in the analysis of Langmuir probe data from low temperature plasmas is examined critically using Doppler spectroscopy of the 468.6 nm He II line. Results for ion temperatures obtained from spectroscopic data for a variety of engineering parameters (discharge voltage, gas pressure and plasma current) will be presented.

High efficiency nebulization for helium inductively coupled plasma mass spectrometry

International Nuclear Information System (INIS)

Jorabchi, Kaveh; McCormick, Ryan; Levine, Jonathan A.; Liu Huiying; Nam, S.-H.; Montaser, Akbar

2006-01-01

A pneumatically-driven, high efficiency nebulizer is explored for helium inductively coupled plasma mass spectrometry. The aerosol characteristics and analyte transport efficiencies of the high efficiency nebulizer for nebulization with helium are measured and compared to the results obtained with argon. Analytical performance indices of the helium inductively coupled plasma mass spectrometry are evaluated in terms of detection limits and precision. The helium inductively coupled plasma mass spectrometry detection limits obtained with the high efficiency nebulizer at 200 μL/min are higher than those achieved with the ultrasonic nebulizer consuming 2 mL/min solution, however, precision is generally better with high efficiency nebulizer (1-4% vs. 3-8% with ultrasonic nebulizer). Detection limits with the high efficiency nebulizer at 200 μL/min solution uptake rate approach those using ultrasonic nebulizer upon efficient desolvation with a heated spray chamber followed by a Peltier-cooled multipass condenser

Testing and analyses of a high temperature thermal barrier for gas-cooled reactors

International Nuclear Information System (INIS)

Black, W.E.; Betts, W.S.; Felten, P.

1979-01-01

A full size, multi-panel section of a thermal barrier system was fabricated from a nickel-base superalloy and a combination of fibrous blanket insulation materials for specific application in a steam cycle gas-cooled nuclear reactor. The 2.4 m square array was representative of the sidewall of the lower core outlet plenum and included coverplates, attachments, seals, and a simulated water-cooled liner. Testing was conducted in a reactor grade, helium-filled chamber at 816 0 C for 100 hours, which established a normal (baseline) condition; 982 0 C for 10 hours, which satisfied an emergency condition; 1093 0 C for 1 hour, which simulated a faulted condition; and 1260 0 C, which was a non-design condition test to demonstrate the temperature overshoot capability of the system. Post-test examination indicated: (1) an acceptable performance by the anti-friction chromium carbide (Cr 3 C 2 ) coating; (2) no significant galling between non-coated surfaces; (3) no distortion of attachment fixtures; (4) predictable coverplate deflection during the design conditions testing (normal, emergency, and faulted); and (5) considerable plastic deformation resulting from the near-incipient melting temperature. (orig.)

POWER CYCLE AND STRESS ANALYSES FOR HIGH TEMPERATURE GAS-COOLED REACTOR

International Nuclear Information System (INIS)

Oh, Chang H; Davis, Cliff; Hawkes, Brian D; Sherman, Steven R

2007-01-01

The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold (1) efficient low cost energy generation and (2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. Many aspects of the NGNP must be researched and developed in order to make recommendations on the final design of the plant. Parameters such as working conditions, cycle components, working fluids, and power conversion unit configurations must be understood. Three configurations of the power conversion unit were demonstrated in this study. A three-shaft design with three turbines and four compressors, a combined cycle with a Brayton top cycle and a Rankine bottoming cycle, and a reheated cycle with three stages of reheat were investigated. An intermediate heat transport loop for transporting process heat to a High Temperature Steam Electrolysis (HTSE) hydrogen production plant was used. Helium, CO2, and a 80% nitrogen, 20% helium mixture (by weight) were studied to determine the best working fluid in terms cycle efficiency and development cost. In each of these configurations the relative component size were estimated for the different working fluids. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the three-shaft and combined cycles were performed to determine the effect of varying conditions in the cycle. This gives some insight into the sensitivity of these cycles to

Testing and analyses of a high temperature duct for gas-cooled reactors

International Nuclear Information System (INIS)

Black, W.E.; Roberge, A.; Felten, P.; Bastien, R.

1979-01-01

A 0.6 scale model of a steam cycle gas-cooled reactor high temperature duct was tested in a closed loop helium facility. The object of the test series was to determine: 1) the thermal effects of gas permeation within the thermal barrier, 2) the plastic deformation of the metallic components, and 3) the thermal performance of the fibrous insulation. A series of tests was performed with thermal cyclings from 100 0 C to 760 0 C at 50 atmospheres until the system thermal performance had stabilized hence enabling predictions for the reactor life. Additional tests were made to assess permeation by deliberately simulating sealing weld failures thereby allowing gas flow by-pass within the primary thermal barrier. After 100 cycles the entire primary structure was found to have performed without structural failure. Due to high pressures exerted by the insulation on the cover plates and a design oversight, the thin seal sheets were unable to expand in an anticipated manner. Local buckling resulted. Pre and post test metallurgical analyses were conducted on the Hastelloy-X structures and reference specimens. The results gave evidence of aging in the form of noticeable changes in room temperature tensile and reduction in area parameters. The Hastelloy-X welds exhibited greater changes in properties due to thermal aging. The antifriction coating (Cr 3 C 2 ) performed well without spallation or excessive wear. (orig.)

Conceptual design of helium experimental loop

International Nuclear Information System (INIS)

Yu Xingfu; Feng Kaiming

2007-01-01

In a future demonstration fusion power station (DEMO), helium is envisaged as coolant for plasma facing components, such as blanket and dive,or. All these components have a very complex geometry, with many parallel cooling channels, involving a complex helium flow distribution. Test blanket modules (TBM) of this concept will under go various tests in the experimental reactor ITER. For the qualification of TBM, it is indispensable to test mock-ups in a helium loop under realistic pressure and temperature profiles, in order to validate design codes, especially regarding mass flow and heat transition processes in narrow cooling channels. Similar testing must be performed for DEMO blanket, currently under development. A Helium Experimental Loop (HELOOP) is planed to be built for TBM tests. The design parameter of temperature, pressure, flow rate is 550 degree C, 10 MPa, l kg/s respectively. In particular, HELOOP is able to: perform full-scale tests of TBM under realistic conditions; test other components of the He-cooling system in ITER; qualify the purification circuit; obtain information for the design of the ITER cooling system. The main requirements and characteristics of the HELOOP facility and a preliminary conceptual design are described in the paper. (authors)

Development of the active magnetic regenerative refrigerator operating between 77 K and 20 K with the conduction cooled high temperature superconducting magnet

Science.gov (United States)

Park, Inmyong; Jeong, Sangkwon

2017-12-01

The experimental investigation of an active magnetic regenerative refrigerator (AMRR) operating between 77 K and 20 K is discussed in this paper, with detailed energy transfer analysis. A multi-layered active magnetic regenerator (AMR) is used, which consists of four different rare earth intermetallic compounds in the form of irregular powder. Numerical simulation confirms that the AMR can attain its target operating temperature range. Magnetic field alternation throughout the AMR is generated by a high temperature superconducting (HTS) magnet. The HTS magnet is cooled by a two stage Gifford-McMahon (GM) cryocooler. Helium gas was employed as a working fluid and its oscillating flow in the AMR is controlled in accordance with the magnetic field variation. The AMR is divided into two stages and each stage has a different mass flow rate as needed to achieve the desired cooling performance. The temperature variation of the AMR during the experiment is monitored by temperature sensors installed inside the AMR. The experimental results show that the AMRR is capable of achieving no-load temperature of 25.4 K while the warm end temperature is 77 K. The performance of the AMRR is analyzed by observing internal temperature variations at cyclic steady state. Furthermore, numerical estimation of the cooling capacity and the temperature variation of the AMR are examined and compared with the experimental results.

Development of VHTR high temperature piping in KHI

International Nuclear Information System (INIS)

Suzuki, Nobuhiro; Takano, Shiro

1981-01-01

The high temperature pipings used for multi-purpose high temperature gas-cooled reactors are the internally insulated pipings for transporting high temperature, high pressure helium at 1000 deg C and 40 kgf/cm 2 , and the influences exerted by their performance as well as safety to the plants are very large. Kawasaki Heavy Industries, Ltd., has engaged in the development of the high temperature pipings for VHTRs for years. In this report, the progress of the development, the test carried out recently and the problems for future are described. KHI manufactured and is constructing a heater and internally insulated helium pipings for the large, high temperature structure testing loop constructed by Japan Atomic Energy Research Institute. The design concept for the high temperature pipings is to separate the temperature boundary and the pressure boundary, therefore, the double walled construction with internal heat insulation was adopted. The requirements for the high temperature pipings are to prevent natural convection, to prevent bypass flow, to minimize radiation heat transfer and to reduce heat leak through insulator supporters. The heat insulator is composed of two layers, metal laminate insulator and fiber insulator of alumina-silica. The present state of development of the high temperature pipings for VHTRs is reported. (Kako, I.)

Oxidation characteristics of the electron beam surface-treated Alloy 617 in high temperature helium environments

International Nuclear Information System (INIS)

Lee, Ho Jung; Sah, Injin; Kim, Donghoon; Kim, Hyunmyung; Jang, Changheui

2015-01-01

The oxidation characteristics of the electron beam surface-treated Alloy 617, which has an Al-rich surface layer, were evaluated in high temperature helium environments. Isothermal oxidation tests were performed in helium (99.999% purity) and VHTR-helium (helium of prototypical VHTR chemistry containing impurities like CO, CO 2 , CH 4 , and H 2 ) environments at 900 °C for up to 1000 h. The surface-treated Alloy 617 showed an initial transient oxidation stage followed by the steady-state oxidation in all test environments. In addition, the steady-state oxidation kinetics of the surface-treated Alloy 617 was 2-order of magnitude lower than that of the as-received Alloy 617 in both helium environments as well as in air. The improvement in oxidation resistance was primarily due to the formation of the protective Al 2 O 3 layer on the surface. The weight gain was larger in the order of air, helium, and VHTR-helium, while the parabolic rate constants (k p ) at steady-state were similar for all test environments. In both helium environments, the oxide structure consisted of the outer transition Al 2 O 3 with a small amount of Cr 2 O 3 and inner columnar structured Al 2 O 3 without an internal oxide. In the VHTR-helium environment, where the impurities were added to helium, the initial transient oxidation increased but the steady state kinetics was not affected

A preliminary definition of the parameters of an experimental natural - uranium, graphite - moderated, helium - cooled power reactor

International Nuclear Information System (INIS)

Baltazar, O.

1978-01-01

A preliminary study of the technical characteristic of an experiment at 32 MWe power with a natural uconium, graphite-moderated, helium cooled reactor is described. The national participation and the use of reactor as an instrument for the technological development of future high temperature gas cooled reactor is considered in the choice of the reactor type. Considerations about nuclear power plants components based in extensive bibliography about similar english GCR reactor is presented. The main thermal, neutronic an static characteristic and in core management of the nuclear fuel is stablished. A simplified scheme of the secondary system and its thermodynamic performance is determined. A scheme of parameters calculation of the reactor type is defined based in the present capacity of calculation developed by Coordenadoria de Engenharia Nuclear and Centro de Processamento de Dados, IEA, Brazil [pt

Brookhaven program to develop a helium-cooled power transmission system

International Nuclear Information System (INIS)

Forsyth, E.B.

1975-01-01

The particular system under design consists of flexible cables installed in a cryogenic enclosure at room temperature and cooled to the range 6 to 9 0 K by supercritical helium, contraction of the cable is accommodated by proper choice of helix angles of the components of the cable. The superconductor is Nb 3 Sn and at the present time the dielectric insulation is still the subject of intensive development. Two good choices appear to be forms of polyethylene and polycarbonate. Sample cables incorporating various dielectrics have been manufactured commercially in lengths of 1500 ft and tested in laboratory cryostats in shorter sections of about 70 ft. A test facility is under construction to evaluate cables and cryogenic components for this type of service, the first refrigerator uses a 350 H.P. screw compressor and three turbo-expander stages. It is hoped to achieve reliability of a very high order. The first three-phase tests will be conducted at 69 kV, although it appears that 230 to 345 kV is the most likely voltage range for future applications. (auth)

Conceptual design of the blanket mechanical attachment for the helium-cooled lithium-lead reactor

International Nuclear Information System (INIS)

Barrera, G.; Branas, B.; Lucas, J.; Doncel, J.; Medrano, M.; Garcia, A.; Giancarli, L.; Ibarra, A.; Li Puma, A.; Maisonnier, D.; Sardain, P.

2008-01-01

The conceptual design of a new type of fusion reactor based on the helium-cooled lithium-lead (HCLL) blanket has been performed within the European Power Plant Conceptual Studies. As part of this activity, a new attachment system suitable for the HCLL blanket modules had to be developed. This attachment is composed of two parts. The first one is the connection between module and the first part of a shield, called high temperature shield, which operates at a temperature around 500 deg. C, close to that of the blanket module. This connection must be made at the lateral walls, in order to avoid openings through the first wall and breeding zone thus avoiding complex design and fabrication issues of the module. The second connection is the one between the high temperature shield and a second shield called low temperature shield, which has a temperature during reactor operation around 150 deg. C. The design of this connection is complex because it must allow the large differential thermal expansion (up to 30 mm) between the two components. Design proposals for both connections are presented, together with the results of finite element mechanical analyses which demonstrate the feasibility to support the blanket and shield modules during normal and accidental operation conditions

High-Temperature Test of 800HT Printed Circuit Heat Exchanger in HELP

International Nuclear Information System (INIS)

Kim, Chan Soo; Hong, Sung-Deok; Kim, Min Hwan; Shim, Jaesool

2014-01-01

Korea Atomic Energy Research Institute has developed high-temperature Printed Circuit Heat Exchangers (PCHE) for a Very High Temperature gas-cooled Reactor and operated a very high temperature Helium Experimental LooP (HELP) to verify the performance of the high temperature heat exchanger at the component level environment. PCHE is one of the candidates for the intermediate heat exchanger in a VHTR, because its design temperature and pressure are larger than any other compact heat exchanger types. High temperature PCHEs in HELP consist of an alloy617 PCHE and an 800HT PCHE. This study presents the high temperature test of an 800HT PCHE in HELP. The experimental data include the pressure drops, the overall heat transfer coefficients, and the surface temperature distributions under various operating conditions. The experimental data are compared with the thermo-hydraulic analysis from COMSOL. In addition, the single channel tests are performed to quantify the friction factor under normal nitrogen and helium inlet conditions. (author)

SAFE AND FAST QUENCH RECOVERY OF LARGE SUPERCONDUCTING SOLENOIDS COOLED BY FORCED TWO-PHASE HELIUM FLOW

International Nuclear Information System (INIS)

Jia, L.X.

1999-01-01

The cryogenic characteristics in energy extraction of the four fifteen-meter-diameter superconducting solenoids of the g-2 magnet are reported in this paper. The energy extraction tests at full-current and half-current of its operating value were deliberately carried out for the quench analyses and evaluation of the cryogenic system. The temperature profiles of each coil mandrel and pressure profiles in its helium cooling tube during the energy extraction are discussed. The low peak temperature and pressure as well as the short recovery time indicated the desirable characteristics of the cryogenic system

A high temperature reactor for ship propulsion

International Nuclear Information System (INIS)

Lobet, P.; Seigel, R.; Thompson, A.C.; Beadnell, R.M.; Beeley, P.A.

2002-01-01

The initial thermal hydraulic and physics design of a high temperature gas cooled reactor for ship propulsion is described. The choice of thermodynamic cycle and thermal power is made to suit the marine application. Several configurations of a Helium cooled, Graphite moderated reactor are then analysed using the WIMS and MONK codes from AEA Technology. Two geometries of fuel elements formed using micro spheres in prismatic blocks, and various arrangements of control rods and poison rods are examined. Reactivity calculations through life are made and a pattern of rod insertion to flatten the flux is proposed and analysed. Thermal hydraulic calculations are made to find maximum fuel temperature under high power with optimized flow distribution. Maximum temperature after loss of flow and temperatures in the reactor vessel are also computed. The temperatures are significantly below the known limits for the type of fuel proposed. It is concluded that the reactor can provide the required power and lifetime between refueling within likely space and weight constraints. (author)

Nuclear power for coexistence with nature, high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Kaneko, Yoshihiko

1996-01-01

Until this century, it is sufficient to aim at the winner of competition in human society to obtain resources, and to entrust waste to natural cleaning action. However, the expansion of social activities has been too fast, and the scale has become too large, consequently, in the next century, the expansion of social activities will be caught by the structure of trilemma that is subjected to the strong restraint and selection from the problems of finite energy and resources and environment preservation. In 21st century, the problems change to those between mankind and nature. Energy supply and population increase, envrionment preservation and human activities, and the matters that human wisdom should bear regarding energy technology are discussed. In Japan, the construction of the high temperature engineering test reactor (HTTR) is in progress. The design of high temperature gas-cooled reactors and their features on the safety are explained. The capability of reducing CO 2 release of high temperature gas-cooled reactors is reported. In future, it is expected that the time of introducing high temperature gas-cooled reactors will come. (K.I.)

Cool-down performance of CICC superconducting coils for the CHMFL

Science.gov (United States)

Xie, Y.; Li, J.; Ouyang, Z. R.

2017-10-01

A hybrid magnet composed of a water-cooled magnet and a superconducting magnet was developed at the High Magnetic Field Laboratory of the Chinese Academy of Sciences. The superconducting coils made of Nb3Sn CICC were cooled by the forced flow of supercritical helium at 4.5 K. The paper presents the cryogenic system framework, and reports the characteristics of the supercritical helium in a cable-in-conduit conductor (CICC), including the friction factor change during the cooling process, the heat transfer coefficient from 4.6 K to 6.8 K, and the helium mass flow rate distribution. After the 23-day cooling process, the temperature reached 4.5 K. The operation process was introduced in the paper.

He-cooled divertor for DEMO. Fabrication technology for tungsten cooling fingers

Energy Technology Data Exchange (ETDEWEB)

Reiser, J.; Norajitra, P.; Widak, V.; Krauss, W. [Forschungszentrum Karlsruhe GmbH (Germany)

2008-07-01

A modular helium-cooled divertor design based on the multi-jet impingement concept (HEMJ) has been developed for the ''post-ITER'' demonstration reactor (DEMO) at the Forschungszentrum Karlsruhe [1, 2]. The main function of the divertor is to keep the plasma free from impurities by catching particles, such as fusion ash and eroded particles from the first wall. From the divertor surface, a maximum heat load of 10 MW/m{sup 2} at least has to be removed. The whole divertor is split up into a number of cassettes (48 according to the latest design studies [3]). Each cassette is cooled separately. The target plates are provided with several cooling fingers to keep the thermal stresses low. Each cooling finger consists of a tungsten tile which is brazed to a thimble-like cap made of a tungsten alloy W-1%La2O3 (WL10) underneath. The thimble has to be connected to the ODS EUROFER steel structure, which is accomplished by brazing again. The tungsten/tungsten brazing is exposed to 1200 C operation temperature while the tungsten/steel brazing joint must withstand 700 C operating temperature. Cooling of the finger is achieved by multi-jet impingement with helium. The inlet temperature of helium is 600 C and rises up to 700 C at the outlet. With this kind of cooling, a mean heat transfer coefficient of 35.000 W/(m{sup 2*}K) can be reached. This compact report will focus on the manufacturing of such a cooling finger unit at FZK. It will cover the machining of the tungsten tile as well as of the thimble and, the brazing of the parts. The major aim of this activity is, on the one hand, to obtain functioning mock-ups with high quality and high reliability, in particular in terms of minimising the surface roughness, cracks, and micro-cracks. On the other hand, effort should also be laid on realising the mass production from economic point of view. (orig.)

The Gas Turbine - Modular Helium Reactor: A Promising Option for Near Term Deployment

International Nuclear Information System (INIS)

LaBar, Malcolm P.

2002-01-01

The Gas Turbine - Modular Helium Reactor (GT-MHR) is an advanced nuclear power system that offers unparalleled safety, high thermal efficiency, environmental advantages, and competitive electricity generation costs. The GT-MHR module couples a gas-cooled modular helium reactor (MHR) with a high efficiency modular Brayton cycle gas turbine (GT) energy conversion system. The reactor and power conversion systems are located in a below grade concrete silo that provides protection against sabotage. The GT-MHR safety is achieved through a combination of inherent safety characteristics and design selections that take maximum advantage of the gas-cooled reactor coated particle fuel, helium coolant and graphite moderator. The GT-MHR is projected to be economically competitive with alternative electricity generation technologies due to the high operating temperature of the gas-cooled reactor, high thermal efficiency of the Brayton cycle power conversion system, high fuel burnup (>100,000 MWd/MT), and low operation and maintenance requirements. (author)

Optimization of advanced high-temperature Brayton cycles with multiple reheat stages

International Nuclear Information System (INIS)

Haihua Zhao; Per F Peterson

2005-01-01

Full text of publication follows: This paper presents an overview and a few point designs for multiple-reheat Brayton cycle power conversion systems using high temperature molten salts (or liquid metals). All designs are derived from the General Atomics GT-MHR power conversion unit (PCU). The GT-MHR PCU is currently the only closed helium cycle system that has undergone detailed engineering design analysis, and that has turbomachinery which is sufficiently large to extrapolate to a >1000 MW(e) multiple reheat gas cycle power conversion system. Analysis shows that, with relatively small engineering modifications, multiple GT-MHR PCU's can be connected together to create a power conversion system in the >1000 MW(e) class. The resulting power conversion system is quite compact, and results in what is likely the minimum gas duct volume possible for a multiple-reheat system. To realize this, compact offset fin plate type liquid-to-gas heat exchangers (power densities from 10 to 120 MW/m 3 ) are needed. Both metal and non-metal heat exchangers are being investigated for high-temperature, gas-cooled reactors for temperatures to 1000 deg. C. Recent high temperature heat exchanger studies for nuclear hydrogen production has suggested that carbon-coated composite materials such as liquid silicon infiltrated chopped fiber carbon-carbon preformed material potentially could be used to fabricate plate fin heat exchangers with reasonable price. Different fluids such as helium, nitrogen and helium mixture, and supercritical CO 2 are compared for these multiple reheat Brayton cycles. Nitrogen and helium mixture cycle need about 40% more total PCU volume than helium cycle while keeping the same net cycle efficiency. Supercritical CO 2 needs very high pressure to optimize. Due to relatively detailed design for components such as heat exchangers, turbomachinery, and duct system, relatively accurate total pressure loss can be obtained, which results in more credible net efficiency

Development status on hydrogen production technology using high-temperature gas-cooled reactor at JAEA, Japan

International Nuclear Information System (INIS)

Shiozawa, Shusaku; Ogawa, Masuro; Hino, Ryutaro

2006-01-01

The high-temperature gas-cooled reactor (HTGR), which is graphite-moderated and helium-cooled, is attractive due to its unique capability of producing high temperature helium gas and its fully inherent reactor safety. In particular, hydrogen production using the nuclear heat from HTGR (up to 900 deg. C) offers one of the most promising technological solutions to curb the rising level of CO 2 emission and resulting risk of climate change. The interests in HTGR as an advanced nuclear power source for the next generation reactor, therefore, continue to rise. This is represented by the Japanese HTTR (High-Temperature Engineering Test Reactor) Project and the Chinese HTR-10 Project, followed by the international Generation IV development program, US nuclear hydrogen initiative program, EU innovative HTR technology development program, etc. To enhance nuclear energy application to heat process industries, the Japan Atomic Energy Agency (JAEA) has continued extensive efforts for development of hydrogen production system using the nuclear heat from HTGR in the framework of the HTTR Project. The HTTR Project has the objectives of establishing both HTGR technology and heat utilization technology. Using the HTTR constructed at the Oarai Research and Development Center of JAEA, reactor performance and safety demonstration tests have been conducted as planned. The reactor outlet temperature of 950 deg. C was successfully achieved in April 2004. For hydrogen production as heat utilization technology, R and D on thermo-chemical water splitting by the 'Iodine-Sulfur process' (IS process) has been conducted step by step. Proof of the basic IS process was made in 1997 on a lab-scale of hydrogen production of 1 L/h. In 2004, one-week continuous operation of the IS process was successfully demonstrated using a bench-scale apparatus with hydrogen production rate of 31 L/h. Further test using a pilot scale facility with greater hydrogen production rate of 10 - 30 m 3 /h is planned as

Low temperature calorimetry and transmission electron microscopy of helium bubbles in Cu

International Nuclear Information System (INIS)

Syskakis, E.

1985-08-01

Helium has been introduced into 100 μm thick pure Cu specimens by implantation of α-particles at T = 300 K. Post-implantation annealing of the specimens at high temperatures caused helium to precipitate into bubbles. We have measured the low-temperature heat capacity of helium confined in bubbles of average radius of less than 100 A. The size of the bubbles was obtained by transmission electron microscope investigations. We have observed that helium liquifies at low temperatures and undergoes the transition to the superfluid state in bubbles of average radius larger than 35 A. The confining geometry of bubbles is new and possesses unique features for investigations of confined helium. It provides the possibility to study properties of extremely small, spherical, completely isolated Bose ''particles'' consisting of 10 4 helium atoms each. Furthermore, as we show, it can be known with better accuracy than formerly investigated confining geometries. (orig./BHO)

Modeling and Simulation of the Sulfur-Iodine Process Coupled to a Very High-Temperature Gas-Cooled Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Hydrogen produced from water using nuclear energy will avoid both the use of fossil fuel and CO{sub 2} emission presumed to be the dominant reason for global warming. A thermo-chemical sulfur-iodine (SI) process coupled to a Very High Temperature Gas-Cooled Reactor(VHTR) is one of the most prospective hydrogen production methods that split water using nuclear energy because the SI process is suitable for large-scale hydrogen production without CO{sub 2} emission. The dynamic simulation code to evaluate the start-up behavior of the chemical reactors placed on the secondary helium loop of the SI process has been developed and partially verified using the steady state values obtained from the Aspen Plus{sup TM} Code simulation. As the start-up dynamic simulation results of the SI process coupled to the IHX, which is one of components in the VHTR system, it is expected that the integrated secondary helium loop of the SI process can be successfully and safely approach the steady state condition.

High power density reactors based on direct cooled particle beds

International Nuclear Information System (INIS)

Powell, J.R.; Horn, F.L.

1985-01-01

Reactors based on direct cooled HTGR type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out long the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBR's) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed. 12 figs

Annual report of the Division of High Temperature Engineering

International Nuclear Information System (INIS)

1982-10-01

Research activities conducted in the Division of High Temperature Engineering during fiscal 1981 are described. R and D works of our division are mainly related to a multi-purpose very high-temperature gas-cooled reactor (VHTR) and a fusion reactor. This report deals with the main results obtained on material test, development of computer codes, heat transfer, fluid-dynamics, structural mechanics and the construction of an M + A (Mother and Adapter) section of a HENDEL (Helium Engineering Demonstration Loop) as well. (author)

Hot helium flow test facility summary report

International Nuclear Information System (INIS)

1980-06-01

This report summarizes the results of a study conducted to assess the feasibility and cost of modifying an existing circulator test facility (CTF) at General Atomic Company (GA). The CTF originally was built to test the Delmarva Power and Light Co. steam-driven circulator. This circulator, as modified, could provide a source of hot, pressurized helium for high-temperature gas-cooled reactor (HTGR) and gas-cooled fast breeder reactor (GCFR) component testing. To achieve this purpose, a high-temperature impeller would be installed on the existing machine. The projected range of tests which could be conducted for the project is also presented, along with corresponding cost considerations

Cooling by mixing of helium isotopes

International Nuclear Information System (INIS)

Hansen, O.P.; Olsen, M.; Rasmussen, F.B.

1975-01-01

The principles of the helium dilution refrigerator are outlined. The lowest temperature attained with a continuously operated dilution refrigerator was about 10 mK, and 5 mK for a limited period when the supply of concentrated 3 He to the mixing chamber was interrupted. (R.S.)

Experimental investigations of flow distribution in coolant system of Helium-Cooled-Pebble-Bed Test Blanket Module

Energy Technology Data Exchange (ETDEWEB)

Ilić, M.; Schlindwein, G., E-mail: georg.schlindwein@kit.edu; Meyder, R.; Kuhn, T.; Albrecht, O.; Zinn, K.

2016-02-15

Highlights: • Experimental investigations of flow distribution in HCPB TBM are presented. • Flow rates in channels close to the first wall are lower than nominal ones. • Flow distribution in central chambers of manifold 2 is close to the nominal one. • Flow distribution in the whole manifold 3 agrees well with the nominal one. - Abstract: This paper deals with investigations of flow distribution in the coolant system of the Helium-Cooled-Pebble-Bed Test Blanket Module (HCPB TBM) for ITER. The investigations have been performed by manufacturing and testing of an experimental facility named GRICAMAN. The facility involves the upper poloidal half of HCPB TBM bounded at outlets of the first wall channels, at outlet of by-pass pipe and at outlets of cooling channels in breeding units. In this way, the focus is placed on the flow distribution in two mid manifolds of the 4-manifold system: (i) manifold 2 to which outlets of the first wall channels and inlet of by-pass pipe are attached and (ii) manifold 3 which supplies channels in breeding units with helium coolant. These two manifolds are connected with cooling channels in vertical/horizontal grids and caps. The experimental facility has been built keeping the internal structure of manifold 2 and manifold 3 exactly as designed in HCPB TBM. The cooling channels in stiffening grids, caps and breeding units are substituted by so-called equivalent channels which provide the same hydraulic resistance and inlet/outlet conditions, but have significantly simpler geometry than the real channels. Using the conditions of flow similarity, the air pressurized at 0.3 MPa and at ambient temperature has been used as working fluid instead of HCPB TBM helium coolant at 8 MPa and an average temperature of 370 °C. The flow distribution has been determined by flow rate measurements at each of 28 equivalent channels, while the pressure distribution has been obtained measuring differential pressure at more than 250 positions. The

A novel nuclear combined power and cooling system integrating high temperature gas-cooled reactor with ammonia–water cycle

International Nuclear Information System (INIS)

Luo, Chending; Zhao, Fuqiang; Zhang, Na

2014-01-01

Highlights: • We propose a novel nuclear ammonia–water power and cooling cogeneration system. • The high temperature reactor is inherently safe, with exhaust heat fully recovered. • The thermal performances are improved compared with nuclear combined cycle. • The base case attains an energy efficiency of 69.9% and exergy efficiency of 72.5%. • Energy conservation and emission reduction are achieved in this cogeneration way. - Abstract: A nuclear ammonia–water power and refrigeration cogeneration system (NAPR) has been proposed and analyzed in this paper. It consists of a closed high temperature gas-cooled reactor (HTGR) topping Brayton cycle and a modified ammonia water power/refrigeration combined bottoming cycle (APR). The HTGR is an inherently safe reactor, and thus could be stable, flexible and suitable for various energy supply situation, and its exhaust heat is fully recovered by the mixture of ammonia and water in the bottoming cycle. To reduce exergy losses and enhance outputs, the ammonia concentrations of the bottoming cycle working fluid are optimized in both power and refrigeration processes. With the HTGR of 200 MW thermal capacity and 900 °C/70 bar reactor-core-outlet helium, the system achieves 88.8 MW net electrical output and 9.27 MW refrigeration capacity, and also attains an energy efficiency of 69.9% and exergy efficiency of 72.5%, which are higher by 5.3%-points and 2.6%-points as compared with the nuclear combined cycle (NCC, like a conventional gas/steam power-only combined cycle while the topping cycle is a closed HTGR Brayton cycle) with the same nuclear energy input. Compared with conventional separate power and refrigeration generation systems, the fossil fuel saving (based on CH 4 ) and CO 2 emission reduction of base-case NAPR could reach ∼9.66 × 10 4 t/y and ∼26.6 × 10 4 t/y, respectively. The system integration accomplishes the safe and high-efficiency utilization of nuclear energy by power and refrigeration

Post irradiation characterization of beryllium and beryllides after high temperature irradiation up to 3000 appm helium production in HIDOBE-01

Energy Technology Data Exchange (ETDEWEB)

Fedorov, A.V., E-mail: fedorov@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, Petten, 1755 ZG (Netherlands); Til, S. van; Stijkel, M.P. [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, Petten, 1755 ZG (Netherlands); Nakamichi, M. [Japan Atomic Energy Agency, Rokkasho (Japan); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/ Josep Pla, n° 2, Torres Diagonal Litoral, Edificio B3, Barcelona 08019 (Spain)

2016-01-15

Titanium beryllides are considered as advanced candidate material for neutron multiplier for the helium cooled pebble bed (HCPB) and/or the water cooled ceramic breeder (WCCB) breeder blankets. In the HIDOBE-01 (HIgh DOse irradiation of BEryllium) experiment, beryllium and beryllide pellets with 5 at% and 7 at% Ti are irradiated at four different target temperatures (T{sub irr}): 425 °C, 525 °C, 650 °C and 750 °C up to the dose corresponding to 3000 appm He production in beryllium. The pellets were supplied by JAEA. During post irradiation examinations the critical properties of volumetric swelling and tritium retention were studied. Both titanium beryllide grades show significantly less swelling than the beryllium grade, with the difference increasing with the irradiation temperature. The irradiation induced swelling was studied by using direct dimensions. Both beryllide grades showed much less swelling as compare to the reference beryllium grade. Densities of the grades were studied by Archimedean immersion and by He-pycnometry, giving indications of porosity formation. While both beryllide grades show no significant reduction in density at all irradiation temperatures, the beryllium density falls steeply at higher T{sub irr}. Finally, the tritium release and retention were studied by temperature programmed desorption (TPD). Beryllium shows the same strong tritium retention as earlier observed in studies on beryllium pebbles, while the tritium inventory of the beryllides is significantly less, already at the lowest T{sub irr} of 425 °C.

Helium refrigeration system for BNL colliding beam accelerator

International Nuclear Information System (INIS)

Brown, D.P.; Farah, Y.; Gibbs, R.J.; Schlafke, A.P.; Schneider, W.J.; Sondericker, J.H.; Wu, K.C.

1983-01-01

A Helium Refrigeration System which will supply the cooling required for the Colliding Beam Accelerator at Brookhaven National Laboratory is under construction. Testing of the compressor system is scheduled for late 1983 and will be followed by refrigerator acceptance tests in 1984. The refrigerator has a design capacity of 24.8 kW at a temperature level near 4K while simultaneously producing 55 kW for heat shield loads at 55K. When completed, the helium refrigerator will be the world's largest. Twenty-five oil-injected screw compressors with an installed total of 23,250 horsepower will supply the gas required. One of the unique features of the cycle is the application of three centrifugal compressors used at liquid helium temperature to produce the low temperatures (2.5K) and high flow rates (4154 g/s) required for this service

Thermal-hydraulics of helium cooled First Wall channels and scoping investigations on performance improvement by application of ribs and mixing devices

Energy Technology Data Exchange (ETDEWEB)

Arbeiter, Frederik, E-mail: frederik.arbeiter@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, Christian [EUROfusion – Programme Management Unit, Garching (Germany); Chen, Yuming; Ilić, Milica; Schwab, Florian [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Sieglin, Bernhard [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Wenninger, Ronald [EUROfusion – Programme Management Unit, Garching (Germany)

2016-11-01

Highlights: • Existing first wall designs and expected plasma heat loads are reviewed. • Heat transfer enhancement methods are investigated by CFD. • The results for heat transfer and friction are given, compared and explained. • Relations for needed pumping power and gained thermal heat are shown. • A range for the maximum permissible heat loads from the plasma is estimated. - Abstract: The first wall (FW) of DEMO is a component with high thermal loads. The cooling of the FW has to comply with the material's upper and lower temperature limits and requirements from stress assessment, like low temperature gradients. Also, the cooling has to be integrated into the balance-of-plant, in a sense to deliver exergy to the power cycle and require a limited pumping power for coolant circulation. This paper deals with the basics of FW cooling and proposes optimization approaches. The effectiveness of several heat transfer enhancement techniques is investigated for the use in helium cooled FW designs for DEMO. Among these are wall-mounted ribs, large scale mixing devices and modified hydraulic diameter. Their performance is assessed by computational fluid dynamics (CFD), and heat transfer coefficients and pressure drop are compared. Based on the results, an extrapolation to high heat fluxes is tried to estimate the higher limits of cooling capabilities.

Cryogenic filter method produces super-pure helium and helium isotopes

Science.gov (United States)

Hildebrandt, A. F.

1964-01-01

Helium is purified when cooled in a low pressure environment until it becomes superfluid. The liquid helium is then filtered through iron oxide particles. Heating, cooling and filtering processes continue until the purified liquid helium is heated to a gas.

Studies on divertor effects by means of the Doublet-III high-temperature plasma device

International Nuclear Information System (INIS)

Shimada, Michiya

1982-12-01

The diverter action on impurity removal, helium ash compression and radiative cooling was studied in Doublet-3, placing emphasis on the applicability to reacting plasma grade devices such as Intor. The following principal results were obtained with a single-null poloidal diverter without the diverter chamber and the diverter throat (referred to as ''open diverter''), and the diverter coils being installed outside the vacuum chamber. The diverter reduced metallic impurities in the central plasma volume, carbon influx and radiation loss, and changed a typically peaked radiation power profile to a hollow profile. In helium-seeded diverter discharge, helium gas pressure near the diverter rose with the increase of main plasma density, and the pressure was high enough to demonstrate the possibility of helium ash exhaust in a diverted tokamak. The radiation power in the diverter volume significantly increased with the increasing main plasma density to as much as 50 % of the input ohmic power. The remote radiation cooling reduced the thermal load on the diverter plate, and the electron temperature near the diverter plate was cooled down. The source of this remote radiative cooling power was the mixture of line radiation of hydrogen neutral and oxygen. (Kako, I.)

Gas cooled thermal reactors with high temperatures (VHTR)

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.; Vasile, A.

2014-01-01

VHTR is one of the 6 concepts retained for the 4. generation of nuclear reactors, it is an upgraded version of the HTR-type reactor (High Temperature Reactors). 5 HTR reactors were operated in the world in the eighties, now 2 experimental HTR are working in China and Japan and 2 HTR with an output power of 100 MWe are being built in China. The purpose of the VHTR is to provide an helium at very high temperatures around 1000 Celsius degrees that could be used directly in a thermochemical way to produce hydrogen for instance. HTR reactors are interesting in terms of safety but it does not optimise the consumption of uranium and the production of wastes. This article presents a brief historical account of HTR-type reactors and their main design and safety features. The possibility of using HTR to burn plutonium is also presented as well as the possibility of closing the fuel cycle and of using thorium-uranium fuel. (A.C.)

Photoassociation of cold metastable helium atoms

NARCIS (Netherlands)

Woestenenk, G.R.

2001-01-01

During the last decades the study of cold atoms has grown in a great measure. Research in this field has been made possible due to the development of laser cooling and trapping techniques. We use laser cooling to cool helium atoms down to a temperature of 1 mK and we are able to

Synergistic effect of helium and hydrogen for bubble swelling in reduced-activation ferritic/martensitic steel under sequential helium and hydrogen irradiation at different temperatures

Energy Technology Data Exchange (ETDEWEB)

Hu, Wenhui [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Chen, Jihong; Luo, Fengfeng; Li, Tiecheng [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Ren, Yaoyao [Center for Electron Microscopy, Wuhan University, Wuhan 430072 (China); Suo, Jinping; Yang, Feng [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

2014-04-15

Highlights: • Helium/hydrogen synergistic effect can increase irradiation swelling of RAFM steel. • Hydrogen can be trapped to the outer surface of helium bubbles. • Too large a helium bubble can become movable. • Point defects would become mobile and annihilate at dislocations at high temperature. • The peak swelling temperature for RAFM steel is 450 °C. - Abstract: In order to investigate the synergistic effect of helium and hydrogen on swelling in reduced-activation ferritic/martensitic (RAFM) steel, specimens were separately irradiated by single He{sup +} beam and sequential He{sup +} and H{sup +} beams at different temperatures from 250 to 650 °C. Transmission electron microscope observation showed that implantation of hydrogen into the specimens pre-irradiated by helium can result in obvious enhancement of bubble size and swelling rate which can be regarded as a consequence of hydrogen being trapped by helium bubbles. But when temperature increased, Ostwald ripening mechanism would become dominant, besides, too large a bubble could become mobile and swallow many tiny bubbles on their way moving, reducing bubble number density. And these effects were most remarkable at 450 °C which was the peak bubble swelling temperature for RAMF steel. When temperature was high enough, say above 450, point defects would become mobile and annihilate at dislocations or surface. As a consequence, helium could no longer effectively diffuse and clustering in materials and bubble formation was suppressed. When temperature was above 500, helium bubbles would become unstable and decompose or migrate out of surface. Finally no bubble was observed at 650 °C.

Synergistic effect of helium and hydrogen for bubble swelling in reduced-activation ferritic/martensitic steel under sequential helium and hydrogen irradiation at different temperatures

International Nuclear Information System (INIS)

Hu, Wenhui; Guo, Liping; Chen, Jihong; Luo, Fengfeng; Li, Tiecheng; Ren, Yaoyao; Suo, Jinping; Yang, Feng

2014-01-01

Highlights: • Helium/hydrogen synergistic effect can increase irradiation swelling of RAFM steel. • Hydrogen can be trapped to the outer surface of helium bubbles. • Too large a helium bubble can become movable. • Point defects would become mobile and annihilate at dislocations at high temperature. • The peak swelling temperature for RAFM steel is 450 °C. - Abstract: In order to investigate the synergistic effect of helium and hydrogen on swelling in reduced-activation ferritic/martensitic (RAFM) steel, specimens were separately irradiated by single He + beam and sequential He + and H + beams at different temperatures from 250 to 650 °C. Transmission electron microscope observation showed that implantation of hydrogen into the specimens pre-irradiated by helium can result in obvious enhancement of bubble size and swelling rate which can be regarded as a consequence of hydrogen being trapped by helium bubbles. But when temperature increased, Ostwald ripening mechanism would become dominant, besides, too large a bubble could become mobile and swallow many tiny bubbles on their way moving, reducing bubble number density. And these effects were most remarkable at 450 °C which was the peak bubble swelling temperature for RAMF steel. When temperature was high enough, say above 450, point defects would become mobile and annihilate at dislocations or surface. As a consequence, helium could no longer effectively diffuse and clustering in materials and bubble formation was suppressed. When temperature was above 500, helium bubbles would become unstable and decompose or migrate out of surface. Finally no bubble was observed at 650 °C

Mixed helium-3 - helium-4 calorimeter. Very low temperature calorimetry; Calorimetre mixte a helium-3 et helium-4. Calorimetrie a tres basse temperature

Energy Technology Data Exchange (ETDEWEB)

Testard, O [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1966-06-01

A description is given of a double-racket calorimeter using helium-4 and helium-3 as the cryogenic fluids and making it possible to vary the temperature continuously from 0.35 K to 4.2 K. By using an electric thermal regulator together with liquid hydrogen it is possible to extend this range up to about 30 K. In the second part, a review is made of the various, methods available for measuring specific heats. The method actually used in the apparatus previously described is described in detail. The difficulties arising from the use of an exchange gas for the thermal contact have been solved by the use of adsorption pumps. (author) [French] On decrit un calorimetre a double enceinte utilisant comme fluide cryogenique l'helium-4 et l'helium-3 et permettant de varier continuement la temperature de 0,35 K a 4,2 K. L'utilisation d'un regulateur thermique electrique ainsi que celle d'hydrogene, liquide permettent d'etendre cette gamme jusqu'a 30 K environ. Dans une deuxieme partie, on passe en revue les diverses methodes de mesure des chaleurs specifiques. La methode concrete utilisee dans l'appareil precedemment decrit est exposee en detail. Les difficultes inherentes a l'utilisation de gaz d'echange comme agent de contact thermique ont ete levees par la mise en oeuvre de pompes a adsorbant. (auteur)

Biennial report of the Department of High Temperature Engineering

International Nuclear Information System (INIS)

1984-10-01

Research activities conducted in the Department of High Temperature Engineering during fiscal 1982 and 1983 are described. Research and development works of the department are mainly related to a multipurpose very high-temperature gas-cooled reactor (VHTR) and a fusion reactor. This report deals with the main results obtained on material test, heat transfer, fluid-dynamics, structural mechanics, development of computer codes and operation of an M + A (Mother and Adapter) section and a T 1 test section of the HENDEL (Helium Engineering Demonstration Loop). (author)

Investigation of thermal transfers in super-fluid helium in porous media

International Nuclear Information System (INIS)

Allain, H.

2009-10-01

Particle accelerators are requiring increased magnetic fields for which niobium tin superconducting magnets are considered. This entails electric insulation and cooling problems. Porous ceramic insulations are potential candidates for cable insulation. As they are permeable to helium, they could allow a direct cooling by super-fluid helium. Therefore, this research thesis deals with the investigation of thermal transfers in superfluid helium in porous media. After a description of an accelerator's superconducting magnet, of its thermodynamics and its various cooling modes, the author describes the physical properties of super-fluid helium, its peculiarities with respect to conventional fluids as well as its different phases (fluid and super-fluid), its dynamics under different regimes (the Landau regime which is similar to the laminar regime for a conventional fluid, and the Gorter-Mellink regime which is the super-fluid turbulent regime). He determines the macroscopic equations governing the He II dynamics in porous media by applying the volume averaging method developed by Whitaker. Theoretical results are validated by comparison with a numerical analysis performed with a numerical code. Then, the author presents the various experimental setups which have been developed for the measurement of the intrinsic permeability, one at room temperature and another at high temperature. Experimental results are discussed, notably with respect to pore size and porosity

Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

Energy Technology Data Exchange (ETDEWEB)

Ingersoll, D.T.

2004-07-29

A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the

Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

International Nuclear Information System (INIS)

Ingersoll, D.T.

2004-01-01

A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the

Severe water ingress accident analysis for a Modular High Temperature Gas Cooled Reactor

International Nuclear Information System (INIS)

Zhang Zuoyi; Scherer, Winfried

1997-01-01

This paper analyzes the severe water ingress accidents in the SIEMENS 200MW Modular High Temperature Gas Cooled Reactor (HTR-Module) under the assumption of no active safety protection systems in order to find the safety margin of the current HTR-Module design. A water, steam and helium multi-phase cavity model is originally developed and implemented in the DSNP simulation system. The developed DSNP system is used to simulate the primary circuit of HTR-Module power plant. The comparisons of the models with the TINTE calculations validate the current simulation. After analyzing the effects of blower separation on water droplets, the wall heat storage, etc., it is found that the maximum H 2 O density increase rate in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduces the impulse of the H 2 O in the reactor core. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600degC was not reached in any case. (author)

Data on loss of off-site electric power simulation tests of the high temperature engineering test reactor

International Nuclear Information System (INIS)

Takeda, Takeshi; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

2002-07-01

The high temperature engineering test reactor (HTTR), the first high temperature gas-cooled reactor (HTGR) in Japan, achieved the first full power of 30 MW on December 7 in 2001. In the rise-to-power test of the HTTR, simulation tests on loss of off-site electric power from 15 and 30 MW operations were carried out by manual shutdown of off-site electric power. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, flow rates of helium and water decreased to the scram points. To shut down the reactor safely, the subcriticality should be kept by the insertion of control rods and the auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components. About 50 s later from the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. This report describes sequences of dynamic components and transient behaviors of the reactor and its cooling system during the simulation tests from 15 and 30 MW operations. (author)

On natural circulation in High Temperature Gas-Cooled Reactors and pebble bed reactors for different flow regimes and various coolant gases

International Nuclear Information System (INIS)

Melesed'Hospital, G.

1983-01-01

The use of CO 2 or N 2 (heavy gas) instead of helium during natural circulation leads to improved performance in both High Temperature Gas-Cooled Reactors (HTGR) and in Pebble Bed Reactors (PBR). For instance, the coolant temperature rise corresponding to a coolant pressure level and a rate of afterheat removal could be only 18% with CO 2 as compared to He, for laminar flow in HTGR; this value would be 40% in PBR. There is less difference between HTGR and PBR for turbulent flows; CO 2 is found to be always better than N 2 . These types of results derived from relationships between coolant properties, coolant flow, temperature rise, pressure, afterheat levels and core geometry, are obtained for HTGR and PBR for various flow regimes, both within the core and in the primary loop

A prestressed concrete pressure vessel for helium high temperature reactor system

International Nuclear Information System (INIS)

Horner, R.M.W.; Hodzic, A.

1976-01-01

A novel prestressed concrete pressure vessel has been developed to provide the primary containment for a fully integrated system comprising a high temperature nuclear reactor, three horizontally mounted helium turbines, associated heat exchangers and inter-connecting ducts. The design and analysis of the pressure vessel is described. Factors affecting the final choice of layout are discussed, and earlier development work seeking to resolve the conflicting requirements of the structural, mechanical, and system engineers outlined. Proposals to increase the present output of about 1000 MW of electrical power to over 3000 MW, by incorporating four turbines in a single pressure vessel are presented. (author)

Seismic test of high temperature piping for HTGR

International Nuclear Information System (INIS)

Kobatake, Kiyokazu; Midoriyama, Shigeru; Ooka, Yuzi; Suzuki, Michiaki; Katsuki, Taketsugu

1983-01-01

Since the high temperature pipings for the high temperature gas-cooled reactor contain helium gas at 1000 deg C and 40 kgf/cm 2 , the double-walled pipe type consisting of the external pipe serving as the pressure boundary and the internal pipe with heat insulating structure was adopted. Accordingly, their aseismatic design is one of the important subjects. Recently, for the purpose of grasping the vibration characteristics of these high temperature pipings and obtaining the data required for the aseismatic design, two specimens, that is, a double-walled pipe model and a heat-insulating structure, were made, and the vibration test was carried out on them, using a 30 ton vibration table of Kawasaki Heavy Industries Ltd. In the high temperature pipings of the primary cooling system for the multi-purpose, high temperature gas-cooled experimental reactor, the external pipes of 32 B bore as the pressure boundary and the internal pipes of 26 B bore with internal heat insulation consisting of double layers of fiber and laminated metal insulators as the temperature boundary were adopted. The testing method and the results are reported. As the spring constant of spacers is larger and clearance is smaller, the earthquake wave response of double-walled pipes is smaller, and it is more advantageous. The aseismatic property of the heat insulation structure is sufficient. (Kako, I.)

Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

International Nuclear Information System (INIS)

Marmier, Alain

2012-01-01

This thesis covers 3 fundamental aspects of High Temperature Reactor (HTR) performance: fuel testing under irradiation for maximized safety and sustainability, fuel architecture for improved economy and sustainability, and a novel Balance of Plant concept to enable future high-tech process heat applications with minimized R and D. The HTR concept features important inherent and passive safety characteristics: high thermal inertia and good thermal conductivity of the core; a negative Doppler coefficient; high quality of fuel elements and low power density. These features keep the core temperature within safe boundaries and minimise fission product release, even in case of severe accidents. The Very High Temperature reactor (VHTR) is based on the same safety concept as the initial HTR, but it aims at offering better economy with a higher reactor outlet temperature (and thus efficiency) and a high fuel discharge burn-up (and thus better sustainability). The inherent safety features of HTR have been demonstrated in small pebble-bed reactors in practice, but have to be replicated for reactors with industrially relevant size and power. An increase of the power density (in order to increase the helium coolant outlet temperature) leads to higher fuel temperatures and therefore higher fuel failure probability. The core of a pebble-bed reactor consists of 6 cm diameter spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. These pebbles contain thousands of 1 mm diameter fuel particles baked into a graphite matrix. These fuel particles, in turn, consist of a fuel kernel with successive coatings of pyrocarbon and silicon carbide layers. The coating layers are designed to contain the fission products that build up during operation of the reactor. The feasibility and performance of the fuel requires experimental verification in view of fuel qualification and licensing. For HTR fuel, the required test string comprises amongst others

Numerical evaluation of flow through a prismatic very high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Barros Filho, Jose A.; Santos, Andre A.C.; Navarro, Moyses A.; Ribeiro, Felipe Lopes

2011-01-01

The High-temperature Gas-cooled reactor (HTGR) is a Next Generation Nuclear System that has a good chance to be used as energy generation source in the near future owing to its potential capacity to supply hydrogen without greenhouse gas emission for the future humanity. Recently, improvements in the HTGR design led to the Very High Temperature Reactor (VHTR) concept in which the outlet temperature of the coolant gas reaches to 1000 deg C increasing the efficiency of the hydrogen and electricity generation. Among the core concepts emerging in the VHTR development stands out the prismatic block which uses coated fuel microspheres named TRISO pressed into cylinders and assembled in hexagonal graphite blocks staked to form columns. The graphite blocks contain flow channels around the fuel cylinders for the helium coolant. In this study an analysis is performed using the CFD code CFX 13.0 on a prismatic fuel assembly in order to investigate its thermo-fluid dynamic performance. The simulations were made in a 1/12 fuel element model of the GT-MHR design which was developed by General Atomics. A numerical mesh verification process based on the Grid Convergence Index (GCI) was performed using five progressively refined meshes to assess the numerical uncertainty of the simulation and determine adequate mesh parameters. An analysis was also performed to evaluate different methods to define the inlet and outlet boundary conditions. In this study simulations of models with and without inlet and outlet plena were compared, showing that the presence of the plena offers a more realistic flow distribution. (author)

Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY2003

International Nuclear Information System (INIS)

2005-03-01

The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research Establishment of The Japan Atomic Energy Research Institute (JAERI) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power. Coolant of helium-gas circulates under the pressure of about 4Mpa, and the reactor inlet and outlet temperature are 395degC and 950degC (maximum), respectively coated particle fuel is used as fuel, and the HTTR core is composed of graphite prismatic blocks. The full power operation of 30MW was attained in December, 2001, and then JAERI received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2003 before the high temperature test operation of 950degC. (author)

Gas-cooled fast-breeder reactor. Helium Circulator Test Facility updated design cost estimate

International Nuclear Information System (INIS)

1979-04-01

Costs which are included in the cost estimate are: Titles I, II, and III Architect-Engineering Services; Titles I, II, and III General Atomic Services; site clearing, grading, and excavation; bulk materials and labor of installation; mechanical and electrical equipment with installation; allowance for contractors' overhead, profit, and insurance; escalation on materials and labor; a contingency; and installation of GAC supplied equipment and materials. The total estimated cost of the facility in As Spent Dollars is $27,700,000. Also included is a cost comparison of the updated design and the previous conceptual design. There would be a considerable penalty for the direct-cooled system over the indirect-cooled system due to the excessive cost of the large diameter helium loop piping to an outdoor heat exchanger. The indirect cooled system which utilizes a helium/Dowtherm G heat exchanger and correspondingly smaller and lower pressure piping to its outdoor air cooler proved to be the more economical of the two systems

Design activity of IHI on the experimental multipurpose high temperature gas-cooled reactor

International Nuclear Information System (INIS)

1978-01-01

With conspicuous interest and attention paid by iron and steel manufacturing industries, the development of the multipurpose high temperature gas-cooled reactor, namely the process heat reactor has been energetically discussed in Japan. The experimental multipurpose high temperature gas-cooled reactor, planned by JAERI (the Japan Atomic Energy Research Institute), is now at the end of the adjustment design stage and about to enter the system synthesizing design stage. The design of the JAERI reactor as a pilot plant for process heat reactors that make possible the direct use of the heat, produced in the reactor, for other industrial uses was started in 1969, and has undergone several revisions up to now. The criticality of the JAERI reactor is expected to be realized before 1985 according to the presently published program. IHI has engaged in the developing work of HTGR (high temperature gas-cooled reactor) including VHTR (very high temperature gas-cooled reactor) for over seven years, producing several achievements. IHI has also participated in the JAERI project since 1973 with some other companies concerned in this field. The design activity of IHI in the development of the JAERI reactor is briefly presented in this paper. (auth.)

Operational experience with double acting piston pumps for cryogenic helium and nitrogen

International Nuclear Information System (INIS)

Vosswinkel, F.

1978-01-01

The design of a high efficiency double acting piston pump suitable for pumping liquefied gases at cryogenic temperatures for cable cooling, is reported. The pump has proved flexible, reliable and efficient in operation. The plunger-type pumps can be used for filling cryostats or dewars with liquid helium or nitrogen from a pressure free or pressurized storage vessel, or as circulators for subcooled, saturated and/or supercritical helium in large scale cooling experiments. Flow rates of up to 17 g/s, maximum operating pressure of 600 kPa absolute and maximum differential pressure of approximately 100 kPa are obtained. (UK)

Metaphysics methods development for high temperature gas cooled reactor analysis

International Nuclear Information System (INIS)

Seker, V.; Downar, T. J.

2007-01-01

Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts in the Generation-IV technology road map. Considerable research has been performed on the design and safety analysis of these reactors. However, the calculational tools being used to perform these analyses are not state-of-the-art and are not capable of performing detailed three-dimensional analyses. This paper presents the results of an effort to develop an improved thermal-hydraulic solver for the pebble bed type high temperature gas cooled reactors. The solution method is based on the porous medium approach and the momentum equation including the modified Ergun's resistance model for pebble bed is solved in three-dimensional geometry. The heat transfer in the pebble bed is modeled considering the local thermal non-equilibrium between the solid and gas, which results in two separate energy equations for each medium. The effective thermal conductivity of the pebble-bed can be calculated both from Zehner-Schluender and Robold correlations. Both the fluid flow and the heat transfer are modeled in three dimensional cylindrical coordinates and can be solved in steady-state and time dependent. The spatial discretization is performed using the finite volume method and the theta-method is used in the temporal discretization. A preliminary verification was performed by comparing the results with the experiments conducted at the SANA test facility. This facility is located at the Institute for Safety Research and Reactor Technology (ISR), Julich, Germany. Various experimental cases are modeled and good agreement in the gas and solid temperatures is observed. An on-going effort is to model the control rod ejection scenarios as described in the OECD/NEA/NSC PBMR-400 benchmark problem. In order to perform these analyses PARCS reactor simulator code will be coupled with the new thermal-hydraulic solver. Furthermore, some of the other anticipated accident scenarios in the benchmark

High Temperature Gas-cooled Reactor Projected Markets and Scoping Economics

Energy Technology Data Exchange (ETDEWEB)

Larry Demick

2010-08-01

The NGNP Project has the objective of developing the high temperature gas-cooled reactor (HTGR) technology to supply high temperature process heat to industrial processes as a substitute for burning of fossil fuels, such as natural gas. Applications of the HTGR technology that have been evaluated by the NGNP Project for supply of process heat include supply of electricity, steam and high-temperature gas to a wide range of industrial processes, and production of hydrogen and oxygen for use in petrochemical, refining, coal to liquid fuels, chemical, and fertilizer plants.

The modular high temperature gas cooled reactor

International Nuclear Information System (INIS)

Lutz, D.E.; Lipps, A.J.

1984-01-01

Due to relatively high operating temperatures, the gas-cooled reactor has the potential to serve a wide variety of energy applications. This paper discusses the energy applications which can be served by the modular HTGR, the magnitude of the potential markets, and the HTGR product cost incentives relative to fossil fuel competition. Advantages of the HTGR modular systems are presented along with a description of the design features and performance characteristics of the current reference HTGR modular systems

HTR-2002: Proceedings of the conference on high temperature reactors

International Nuclear Information System (INIS)

2002-01-01

High temperature reactors are considered as future inherently safe and efficient energy sources. The presentations covered all the relevant aspects of the existing HTGRs and/or helium cooled pebble bed reactors. They were sorted into 7 sessions: HTR Projects and Programmes; Fuel and Fuel Cycle; Physics and Neutronics; Thermohydraulic Calculation; Engineering, Design and Applications; Materials and Components; Safety and Licensing

Construction and performance tests of Helium Engineering Demonstration Loop (HENDEL) for VHTR

International Nuclear Information System (INIS)

Hishida, M.; Tanaka, T.; Shimomura, H.; Sanokawa, K.

1984-01-01

A helium engineering demonstration loop (HENDEL) was constructed and operated in JAERI in order to develop the high-temperature key components of an experimental very high temperature gas cooled reactor, like fuel stack, in-core reactor structure, hot gas duct, intermediate heat exchanger. Performance tests as well as demonstration of integrity are carried out with large-size or actual-size models of key components. The key components to be tested in HENDEL are: fuel stack and control rod; core supporting structure, or bottom structure of rector core exposed to direct impingement of high temperature core outlet flow; reactor internal components and structure; high temperature components in heat removal system (primary and secondary cooling systems)

Numerical investigation of heat transfer in high-temperature gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Chen, g.; Anghaie, S. [Univ. of Florida, Gainesville, FL (United States)

1995-09-01

This paper proposes a computational model for analysis of flow and heat transfer in high-temperature gas-cooled reactors. The formulation of the problem is based on using the axisymmetric, thin layer Navier-Stokes equations. A hybrid implicit-explicit method based on finite volume approach is used to numerically solve the governing equations. A fast converging scheme is developed to accelerate the Gauss-Siedel iterative method for problems involving the wall heat flux boundary condition. Several cases are simulated and results of temperature and pressure distribution in the core are presented. Results of a parametric analysis for the assessment of the impact of power density on the convective heat transfer rate and wall temperature are discussed. A comparative analysis is conducted to identify the Nusselt number correlation that best fits the physical conditions of the high-temperature gas-cooled reactors.

A MEASUREMENT OF THE ADIABATIC COOLING INDEX FOR INTERSTELLAR HELIUM PICKUP IONS IN THE INNER HELIOSPHERE

International Nuclear Information System (INIS)

Saul, Lukas; Wurz, Peter; Kallenbach, Reinald

2009-01-01

Interstellar neutral gas enters the inner heliosphere where it is ionized and becomes the pickup ion population of the solar wind. It is often assumed that this population will subsequently cool adiabatically, like an expanding ideal gas due, to the divergent flow of the solar wind. Here, we report the first independent measure of the effective adiabatic cooling index in the inner heliosphere from SOHO CELIAS measurements of singly charged helium taken during times of perpendicular interplanetary magnetic field. We use a simple adiabatic transport model of interstellar pickup helium ions, valid for the upwind region of the inner heliosphere. The time averaged velocity spectrum of helium pickup ions measured by CELIAS/CTOF is fit to this model with a single free parameter which indicates an effective cooling rate with a power-law index of γ = 1.35 ± 0.2. While this average is consistent with the 'ideal-gas' assumption of γ = 1.5, the analysis indicates that such an assumption will not apply in general, and that due to observational constraints further measurements are necessary to constrain the cooling process. Implications are discussed for understanding the transport processes in the inner heliosphere and improving this measurement technique.

Pressure transients analysis of a high-temperature gas-cooled reactor with direct helium turbine cycle

Energy Technology Data Exchange (ETDEWEB)

Dang, M.; Dupont, J. F.; Jacquemoud, P.; Mylonas, R. [Eidgenoessisches Inst. fuer Reaktorforschung, Wuerenlingen (Switzerland)

1981-01-15

The direct coupling of a gas cooled reactor with a closed gas turbine cycle leads to a specific dynamic plant behaviour, which may be summarized as follows: a) any operational transient involving a variation of the core mass flow rate causes a variation of the pressure ratio of the turbomachines and leads unavoidably to pressure and temperature transients in the gas turbine cycle; and b) very severe pressure equalization transients initiated by unlikely events such as the deblading of one or more turbomachines must be taken into account. This behaviour is described and illustrated through results gained from computer analyses performed at the Swiss Federal Institute for Reactor Research (EIR) in Wurenlingen within the scope of the Swiss-German HHT project.

Power Conversion Study for High Temperature Gas-Cooled Reactors

International Nuclear Information System (INIS)

Chang Oh; Richard Moore; Robert Barner

2005-01-01

The Idaho National Laboratory (INL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. There are some technical issues to be resolved before the selection of the final design of the high temperature gas cooled reactor, called as a Next Generation Nuclear Plant (NGNP), which is supposed to be built at the INEEL by year 2017. The technical issues are the selection of the working fluid, direct vs. indirect cycle, power cycle type, the optimized design in terms of a number of intercoolers, and others. In this paper, we investigated a number of working fluids for the power conversion loop, direct versus indirect cycle, the effect of intercoolers, and other thermal hydraulics issues. However, in this paper, we present part of the results we have obtained. HYSYS computer code was used along with a computer model developed using Visual Basic computer language

The design, safety and project development status of the modular high temperature gas-cooled reactor in the United States

International Nuclear Information System (INIS)

Mears, L.D.; Dean, R.A.

1987-01-01

The cooperative government and industry Modular High Temperature Gas-Cooled Reactor (MHTGR) Program in the United States has advanced a 350 MW(t) plant design through the conceptual development stage. The system incorporates an annular core of prismatic fuel elements within a steel pressure vessel connected, in a side-by-side arrangement, by a concentric duct to a second steel vessel containing a steam generator and helium coolant circulator. The reference plant design consists of four reactor modules installed in separate below-grade silos, providing steam to two conventional turbine generators. The nominal net plant output is 540 MW(e). The small reactor system takes unique advantage of the high temperature capability of the refractory coated fuel and the large thermal inertia of the graphite moderator to provide a design capable of withstanding a complete loss of active core cooling without causing excessive core heatup and significant release of fission products from the fuel. Present program activities are concentrated on interactions with the Nuclear Regulatory Commission aimed at obtaining a Licensability Statement. A project initiative to build a prototype plant which would demonstrate the MHTGR-unique licensing process, plant performance, costs and schedule plus establish an industrial infrastructure to proceed with follow-on commercial MHTGR plants by the turn of the century, is being undertaken by the utility/vendor participants (author)

Development of plate-fin heat exchanger for intermediate heat exchanger of high-temperature gas cooled reactor. Fabrication process, high-temperature strength and creep-fatigue life prediction of plate-fin structure made of Hastelloy X

International Nuclear Information System (INIS)

Mizokami, Yorikata; Igari, Toshihide; Nakashima, Keiichi; Kawashima, Fumiko; Sakakibara, Noriyuki; Kishikawa, Ryouji; Tanihira, Masanori

2010-01-01

The helium/helium heat exchanger (i.e., intermediate heat exchanger: IHX) of a high-temperature gas-cooled reactor (HTGR) system with nuclear heat applications is installed between a primary system and a secondary system. IHX is operated at the highest temperature of 950degC and has a high capacity of up to 600 MWt. A plate-fin-type heat exchanger is the most suitable for IHX to improve construction cost. The purpose of this study is to develop an ultrafine plate-fin-type heat exchanger with a finer pitch fin than a conventional technology. In the first step, fabrication conditions of the ultrafine plate fin were optimized by press tests. In the second step, a brazing material was selected from several candidates through brazing tests of rods, and brazing conditions were optimized for plate-fin structures. In the third step, tensile strength, creep rupture, fatigue, and creep-fatigue tests were performed as typical strength tests for plate-fin structures. The obtained data were compared with those of the base metal and plate-fin element fabricated from SUS316. Finally, the accuracy of the creep-fatigue life prediction using both the linear cumulative damage rule and the equivalent homogeneous solid method was confirmed through the evaluation of creep-fatigue test results of plate-fin structures. (author)

Summary of ORNL high-temperature gas-cooled reactor program

International Nuclear Information System (INIS)

Kasten, P.R.

1981-01-01

Oak Ridge National Laboratory (ORNL) efforts on the High-Temperature Gas-Cooled Reactor (HTGR) Program have been on HTGR fuel development, fission product and coolant chemistry, prestressed concrete reactor vessel (PCRV) studies, materials studies, graphite development, reactor physics and shielding studies, application assessments and evaluations and selected component testing

Relative thermoluminescent efficiencies proton/gamma and helium/gamma of peaks of high temperature in TLD-100 dosemeters

International Nuclear Information System (INIS)

Flores M, E.

2007-01-01

The increase of the applications of ion beams in radiotherapy treatments has generated interest in the study of the thermoluminescent materials (TL) that allow to determine the applied doses. A way to quantify the TL response from these materials to ions is by means of the relative thermoluminescent efficiency. In the group of Thermoluminescent dosimetry of the Institute of Physics of the UNAM (IFUNAM) the thermoluminescent response of the TLD-100 dosemeters has been studied, which present a glow curve characteristic with several peaks that correspond to traps and luminescent centers in the material. The stable peaks know each other as 4, 5, 6a, 6b, 7, 8, 9 and 10. The efficiencies should be measured using the response so much to the radiation of interest (in this case protons and helium ions) as the response to gamma radiation. In previous works with ions of low energy taken place in the Pelletron accelerator of the IFUNAM was only measured the TL efficiency for the peak 5 and the total signal. It had not been possible to measure the efficiency of the peaks of high temperature (6a-10) because, for the gamma radiation, the peaks of high temperature show very small signals; however, recently Massillon carries out measures of efficiency TL of peaks of high temperature for ions of intermediate energy using a protocol special of reading and of deconvolution that allows to measure the signals coming from the peaks of high temperature. In this work is implemented this same protocol to complete the study of TL efficiencies at low energy of protons and helium and to determine if the values of efficiency depend on the used reading protocol. For it is reported it measures of the relative efficiency of the peaks of high temperature from the TLD-100 exposed to protons of 1.5 MeV and nuclei of helium of 3 and 7.5 MeV. (Author)

Non-linear Model Predictive Control for cooling strings of superconducting magnets using superfluid helium

CERN Document Server

AUTHOR|(SzGeCERN)673023; Blanco Viñuela, Enrique

In each of eight arcs of the 27 km circumference Large Hadron Collider (LHC), 2.5 km long strings of super-conducting magnets are cooled with superfluid Helium II at 1.9 K. The temperature stabilisation is a challenging control problem due to complex non-linear dynamics of the magnets temperature and presence of multiple operational constraints. Strong nonlinearities and variable dead-times of the dynamics originate at strongly heat-flux dependent effective heat conductivity of superfluid that varies three orders of magnitude over the range of possible operational conditions. In order to improve the temperature stabilisation, a proof of concept on-line economic output-feedback Non-linear Model Predictive Controller (NMPC) is presented in this thesis. The controller is based on a novel complex first-principles distributed parameters numerical model of the temperature dynamics over a 214 m long sub-sector of the LHC that is characterized by very low computational cost of simulation needed in real-time optimizat...

Strength analysis of CARR-CNS with crescent-shape moderator cell and helium sub-cooling jacket covering cell

International Nuclear Information System (INIS)

Yu Qingfeng; Feng Quanke; Kawai Takeshi; Shen Feng; Yuan Luzheng; Cheng Liang

2005-01-01

The new type of the moderator cell was developed for the cold neutron source (CNS) of the China Advanced Research Reactor (CARR) which is now being constructed at the China Institute of Atomic Energy in Beijing. A crescent-shape moderator cell covered by the helium sub-cooling jacket is adopted. The structure of the moderator cell is optimized by the stress FEM analysis. A crescent-shape would help to increase the volume of the moderator cell for fitting it to the four cold neutron guide tubes, even if liquid hydrogen, not liquid deuterium, was used as a cold moderator. The helium sub-cooling jacket covering the moderator cell removes the nuclear heating of the outer shell wall of the cell. It contributes to reduce the void fraction of liquid hydrogen in the outer shell of the moderator cell. Such a type of a moderator cell is suitable for the CNS with higher nuclear heating. The cold helium gas flows down first into the helium sub-cooling jacket and then flows up to the condenser. The theory of the self-regulation suitable to the thermo-siphon type of the CNS is also applicable and validated

Design and analysis of helium Brayton power cycles for HiPER reactor

Energy Technology Data Exchange (ETDEWEB)

Sánchez, Consuelo, E-mail: csanchez@ind.uned.es [Dpto. Ingeniería Energética UNED, Madrid (Spain); Juárez, Rafael; Sanz, Javier [Dpto. Ingeniería Energética UNED, Madrid (Spain); Instituto de Fusión Nuclear/UPM, Madrid (Spain); Perlado, Manuel [Instituto de Fusión Nuclear/UPM, Madrid (Spain)

2013-10-15

Highlights: ► A helium Brayton cycle has been designed integrating the two energy sources of HiPER. ► The Brayton cycle has intercooling stages and a recovery process. ► The low temperature of HiPER heat sources results in low cycle efficiency (35.2%). ► Two inter-cooling stages and a reheating process increases efficiency to over 37%. ► Helium Brayton cycles are to be considered as candidates for HiPER power cycles. -- Abstract: Helium Brayton cycles have been studied as power cycles for both fission and fusion reactors obtaining high thermal efficiency. This paper studies several technological schemes of helium Brayton cycles applied for the HiPER reactor proposal. Since HiPER integrates technologies available at short term, its working conditions results in a very low maximum temperature of the energy sources, something that limits the thermal performance of the cycle. The aim of this work is to analyze the potential of the helium Brayton cycles as power cycles for HiPER. Several helium Brayton cycle configurations have been investigated with the purpose of raising the cycle thermal efficiency under the working conditions of HiPER. The effects of inter-cooling and reheating have specifically been studied. Sensitivity analyses of the key cycle parameters and component performances on the maximum thermal efficiency have also been carried out. The addition of several inter-cooling stages in a helium Brayton cycle has allowed obtaining a maximum thermal efficiency of over 36%, and the inclusion of a reheating process may also yield an added increase of nearly 1 percentage point to reach 37%. These results confirm that helium Brayton cycles are to be considered among the power cycle candidates for HiPER.

Design and analysis of helium Brayton power cycles for HiPER reactor

International Nuclear Information System (INIS)

Sánchez, Consuelo; Juárez, Rafael; Sanz, Javier; Perlado, Manuel

2013-01-01

Highlights: ► A helium Brayton cycle has been designed integrating the two energy sources of HiPER. ► The Brayton cycle has intercooling stages and a recovery process. ► The low temperature of HiPER heat sources results in low cycle efficiency (35.2%). ► Two inter-cooling stages and a reheating process increases efficiency to over 37%. ► Helium Brayton cycles are to be considered as candidates for HiPER power cycles. -- Abstract: Helium Brayton cycles have been studied as power cycles for both fission and fusion reactors obtaining high thermal efficiency. This paper studies several technological schemes of helium Brayton cycles applied for the HiPER reactor proposal. Since HiPER integrates technologies available at short term, its working conditions results in a very low maximum temperature of the energy sources, something that limits the thermal performance of the cycle. The aim of this work is to analyze the potential of the helium Brayton cycles as power cycles for HiPER. Several helium Brayton cycle configurations have been investigated with the purpose of raising the cycle thermal efficiency under the working conditions of HiPER. The effects of inter-cooling and reheating have specifically been studied. Sensitivity analyses of the key cycle parameters and component performances on the maximum thermal efficiency have also been carried out. The addition of several inter-cooling stages in a helium Brayton cycle has allowed obtaining a maximum thermal efficiency of over 36%, and the inclusion of a reheating process may also yield an added increase of nearly 1 percentage point to reach 37%. These results confirm that helium Brayton cycles are to be considered among the power cycle candidates for HiPER

High-Temperature Gas-Cooled Test Reactor Point Design

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

2016-04-01

A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

Quantum statistics and liquid helium 3 - helum 4 mixtures

International Nuclear Information System (INIS)

Cohen, E.G.D.

1979-01-01

The behaviour of liquid helium 3-helium 4 mixtures is considered from the point of view of manifestation of quantum statistics effects in macrophysics. The Boze=Einstein statistics is shown to be of great importance for understanding superfluid helium-4 properties whereas the Fermi-Dirac statistics is of importance for understanding helium-3 properties. Without taking into consideration the interaction between the helium atoms it is impossible to understand the basic properties of liquid helium 33 - helium 4 mixtures at constant pressure. Proposed is a simple model of the liquid helium 3-helium 4 mixture, namely the binary mixture consisting of solid spheres of two types subjecting to the Fermi-Dirac and Bose-Einstein statistics relatively. This model predicts correctly the most surprising peculiarities of phase diagrams of concentration dependence on temperature for helium solutions. In particular, the helium 4 Bose-Einstein statistics is responsible for the phase lamination of helium solutions at low temperatures. It starts in the peculiar critical point. The helium 4 Fermi-Dirac statistics results in incomplete phase lamination close to the absolute zero temperatures, that permits operation of a powerful cooling facility, namely refrigerating machine on helium solution

Influence of the ambient temperature on the cooling efficiency of the high performance cooling device with thermosiphon effect

Science.gov (United States)

Nemec, Patrik; Malcho, Milan

2018-06-01

This work deal with experimental measurement and calculation cooling efficiency of the cooling device working with a heat pipe technology. The referred device in the article is cooling device capable transfer high heat fluxes from electric elements to the surrounding. The work contain description, working principle and construction of cooling device. The main factor affected the dissipation of high heat flux from electronic elements through the cooling device to the surrounding is condenser construction, its capacity and option of heat removal. Experimental part describe the measuring method cooling efficiency of the cooling device depending on ambient temperature in range -20 to 40°C and at heat load of electronic components 750 W. Measured results are compared with results calculation based on physical phenomena of boiling, condensation and natural convection heat transfer.

Computer simulation of multiple stability regions in an internally cooled superconducting conductor and of helium replenishment in a bath-cooled conductor

International Nuclear Information System (INIS)

Turner, L.R.; Shindler, J.

1984-09-01

For upcoming fusion experiments and future fusion reactors, superconducting magnetic have been chosen or considered which employ cooling by pool-boiling HeI, by HeII, and by internally flowing HeI. The choice of conductor and cooling method should be determined in part by the response of the magnet to sudden localized heat pulses of various magnitudes. The paper describes the successful computer simulation of multiple stability in internally cooled conductors, as observed experimentally, using the computer code SSICC. It also describes the modeling of helium replenishment in the cooling channels of a bath-cooled conductor, using the computer code TASS

NUCLEAR CONDENSATE AND HELIUM WHITE DWARFS

Energy Technology Data Exchange (ETDEWEB)

Bedaque, Paulo F.; Berkowitz, Evan [Maryland Center for Fundamental Physics, Department of Physics, University of Maryland, College Park, MD (United States); Cherman, Aleksey, E-mail: bedaque@umd.edu, E-mail: evanb@umd.edu, E-mail: a.cherman@damtp.cam.ac.uk [Department of Applied Mathematics and Theoretical Physics, University of Cambridge, Cambridge CB3 0WA (United Kingdom)

2012-04-10

We consider a high-density region of the helium phase diagram, where the nuclei form a Bose-Einstein condensate rather than a classical plasma or a crystal. Helium in this phase may be present in helium-core white dwarfs. We show that in this regime there is a new gapless quasiparticle not previously noticed, arising when the constraints imposed by gauge symmetry are taken into account. The contribution of this quasiparticle to the specific heat of a white dwarf core turns out to be comparable in a range of temperatures to the contribution from the particle-hole excitations of the degenerate electrons. The specific heat in the condensed phase is two orders of magnitude smaller than in the uncondensed plasma phase, which is the ground state at higher temperatures, and four orders of magnitude smaller than the specific heat that an ion lattice would provide, if formed. Since the specific heat of the core is an important input for setting the rate of cooling of a white dwarf star, it may turn out that such a change in the thermal properties of the cores of helium white dwarfs has observable implications.

NUCLEAR CONDENSATE AND HELIUM WHITE DWARFS

International Nuclear Information System (INIS)

Bedaque, Paulo F.; Berkowitz, Evan; Cherman, Aleksey

2012-01-01

We consider a high-density region of the helium phase diagram, where the nuclei form a Bose-Einstein condensate rather than a classical plasma or a crystal. Helium in this phase may be present in helium-core white dwarfs. We show that in this regime there is a new gapless quasiparticle not previously noticed, arising when the constraints imposed by gauge symmetry are taken into account. The contribution of this quasiparticle to the specific heat of a white dwarf core turns out to be comparable in a range of temperatures to the contribution from the particle-hole excitations of the degenerate electrons. The specific heat in the condensed phase is two orders of magnitude smaller than in the uncondensed plasma phase, which is the ground state at higher temperatures, and four orders of magnitude smaller than the specific heat that an ion lattice would provide, if formed. Since the specific heat of the core is an important input for setting the rate of cooling of a white dwarf star, it may turn out that such a change in the thermal properties of the cores of helium white dwarfs has observable implications.

Potentialities in electronics of new high critical temperature superconductors

International Nuclear Information System (INIS)

Hartemann, P.

1989-01-01

The main electronic applications of superconductors involve the signal processing, the electromagnetic wave detection and the magnetometry. Characteristics of devices based on conventional superconductors cooled by liquid helium are given and the changes induced by incorporating high-temperature superconductors are estimated. After a survey of new superconductor properties, the superconducting devices for analog or digital signal processing are reviewed. The gains predicted for high-temperature superconducting analog devices are considered in greater detail. Different sections deal with the infrared or (sub)millimeter wave detection. The most sensitive apparatuses for magnetic measurements are based on SQUIDs. Features of SQUIDs made of granular high-temperature superconducting material samples (grain boundaries behave as barriers of intrinsic junctions) are discussed [fr

The calculating methods of the release of airborne radionuclides to environment during the normal operation of a module high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Liu Yuanzhong

1993-01-01

The calculations of the release of radionuclides to environment are the basis of environmental impact assessment during the normal operation of a module high temperature gas-cooled reactor of the Institute of Nuclear Energy Technology, Tsinghua University, China. According to the features of the reactor it is pointed out that only five sources of the airborne radioactive materials released to environment are important. They are: (1) the activation of the air in the reactor cavity; (2) the escape from the primary coolant systems; (3) the release of radioactively contaminated helium from storage tanks; (4) the release of radioactively contaminated helium from the gas evacuation system of fuel load and unload system; (5) the leakage of the vapour from water-steam loop. In accordance with five release sources the calculating methods of radionuclides released to environment are worked out respectively and the respective calculating formulas are derived for the normal operation of the reactor

Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

Directory of Open Access Journals (Sweden)

Fic Adam

2015-03-01

Full Text Available Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle, which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle. The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

Creep and fatigue properties of Incoloy 800H in a high-temperature gas-cooled reactor (HTGR) helium environment

International Nuclear Information System (INIS)

Chow, J.G.Y.; Soo, P.; Epel, L.

1978-01-01

A mechanical test program to assess the effects of a simulated HTGR helium environment on the fatigue and creep properties of Incoloy 800H and other primary-circuit metals is described. The emphasis and the objectives of this work are directed toward obtaining information to assess the integrity and safety of an HTGR throughout its service life. The helium test environment selected for study contained 40 μ atm H 2 O, 200 μ atm H 2 , 40 μ atm CO, 10 μ atm CO 2 , and 20 μ atm CH 4 . It is believed that this ''wet'' environment simulates that which could exist in a steam-cycle HTGR containing some leaking steam-generator tubes. A recirculating helium loop operating at about 4 psi in which impurities can be maintained at a constant level, has been constructed to supply the desired environment for fatigue and creep testing

Research and development for high temperature gas cooled reactor in Japan

International Nuclear Information System (INIS)

Taketani, K.

1978-01-01

The paper describes the current status of High Temperature Gas Cooled Reactor research and development work in Japan, with emphasis on the Experimental Very High Temperature Reactor (Exp. VHTR) to be built by Japan Atomic Energy Research Institute (JAERI) before the end of 1985. The necessity of construction of Exp. VHTR was explained from the points of Japanese energy problems and resources

Construction and testing of a double acting bellows liquid helium pump

International Nuclear Information System (INIS)

Burns, W.A.; Green, M.A.; Ross, R.R.; Van Slyke, H.

1980-05-01

The double acting reciprocating bellows liquid helium pump built and tested at the Lawrence Berkeley Laboratory is described. The pump is capable of delivering 50 gs -1 of liquid helium to supply the two-phase cooling sytem for a large superconducting magnet. The pump is driven by a torque motor at room temperature; the reciprocating motion is transmitted to the pump through a shaft which operates between room temperature and 4 0 K. The design details of this liquid helium pump are presented. The helium pump has operated in a helium bath and in pumped forced flow helium circuits. The results of these experimental tests are presented in this report

Cooling-water amounts, temperature, and the environment

International Nuclear Information System (INIS)

Koops, F.B.J.; Donze, M.; Hadderingh, R.H.

1979-01-01

The release of heat from power plants into a water can take place with relative small quantities of cooling water, highly warmed up accordingly, or with large quantities of cooling water slightly warmed up. The utilization of cooling water is bound to certain guidelines established by the authorities. With the intention to protect the environment, the admissable temperatures and warming-up have been strictly limited by the authorities. In the Netherlands, we have presently temporary cooling water guidelines which allow a max. temperature of the cooling water in the cooling cycle of 30 0 C and a maximum admissible temperature rise in the condenser between 7 0 C during summer and 15 0 C during winter. It has also been determined in these requirements how much cooling water at least has to be used to discharge a specified quantity of heat. Plankton, spawn and young fish are dragged with the cooling water. Harm to these organisms can be caused mechanically by pumps, sieves and the condenser or they can be harmed by the temperature rise in the condenser. Investigations showed that mechanical harm to spawn and young fish in the cooling water flow should not be ignored, and that detectable harm to plankton organisms takes place only at water temperatures above 32 0 C. The cooling water consumption can therefore be optimised as follows: The solution of a greater temperature increase and a slightly higher value for the temperature maximum can reduce the cooling water quantity. This reduction of the cooling water quantity reduces the destruction of the fish quantity, which gets into the cooling water system, especially during the summer. If the temperature rise and the temperature itself are not selected too high, the destruction of fish may be reduced without causing serious damage to the plankton. (orig.) [de

Thermal-hydraulic optimization of flexible transfer lines for liquid helium; Thermohydraulische Optimierung flexibler Transferleitungen fuer Fluessighelium

Energy Technology Data Exchange (ETDEWEB)

Dittmar, Nico; Haberstroh, Christoph; Hesse, U. [Technische Univ. Dresden (Germany). Bitzer-Stiftungsprofessur fuer Kaelte-, Kryo- und Kompressorentechnik; Wolfram, M.; Krzyzowski, M.; Raccanelli, A. [CryoVac Gesellschaft fuer Tieftemperaturtechnik mbH und Co. KG, Troisdorf (Germany)

2014-07-01

Cooling systems and applications at very low temperatures are based on the use of liquid helium as cryogenic agent; the normal boiling temperature of helium-4 is 4.2 K. Due to the restricted economic production possibilities and the high energetic expenditure for helium liquefaction an efficient and sustainable handling with the resources is recommended. In university facilities the liquid helium is usually stored in containers and filled into smaller containers for transport using cryogenic transfer lines. This procedure can cause 20% loss by evaporation due to heat input and friction pressure losses. The gaseous helium has to be collected for re-liquefaction. The contribution shows that using systematic measurements an increase of the transfer rate and the efficiency of the helium filling system can be reached by a modified transfer line design.

Safety aspects of forced flow cooldown transients in modular high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Kroeger, P.G.

1992-01-01

During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs) the main Heat Transport System (HTS) and the Shutdown Cooling System (SCS), are assumed to have failed. Decay heat is then removed by the passive Reactor Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This paper uses the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits

Forced flow cooling of ISABELLE dipole magnets

International Nuclear Information System (INIS)

Bamberger, J.A.; Aggus, J.; Brown, D.P.; Kassner, D.A.; Sondericker, J.H.; Strobridge, T.R.

1976-01-01

The superconducting magnets for ISABELLE will use a forced flow supercritical helium cooling system. In order to evaluate this cooling scheme, two individual dipole magnets were first tested in conventional dewars using pool boiling helium. These magnets were then modified for forced flow cooling and retested with the identical magnet coils. The first evaluation test used a l m-long ISA model dipole magnet whose pool boiling performance had been established. The same magnet was then retested with forced flow cooling, energizing it at various operating temperatures until quench occurred. The magnet performance with forced flow cooling was consistent with data from the previous pool boiling tests. The next step in the program was a full-scale ISABELLE dipole ring magnet, 4.25 m long, whose performance was first evaluated with pool boiling. For the forced flow test the magnet was shrunk-fit into an unsplit laminated core encased in a stainless steel cylinder. The high pressure gas is cooled below 4 K by a helium bath which is pumped below atmospheric pressure with an ejector nozzle. The performance of the full-scale dipole magnet in the new configuration with forced flow cooling, showed a 10 percent increase in the attainable maximum current as compared to the pool boiling data

Modular helium reactor for non-electric applications

International Nuclear Information System (INIS)

Shenoy, A.

1997-01-01

The high temperature gas-cooled Modular Helium Reactor (MHR) is an advanced, high efficiency reactor system which can play a vital role in meeting the future energy needs of the world by contributing not only to the generation of electric power, but also the non-electric energy traditionally served by fossil fuels. This paper summarizes work done over 20 years, by several people at General Atomics, how the Modular Helium Reactor can be integrated to provide different non-electric applications during Process Steam/Cogeneration for industrial application, Process Heat for transportation fuel development and Hydrogen Production for various energy applications. The MHR integrates favorably into present petrochemical and primary metal process industries, heavy oil recovery, and future shale oil recovery and synfuel processes. The technical fit of the Process Steam/Cogeneration Modular Helium Reactor (PS/C-MHR) into these processes is excellent, since it can supply the required quantity and high quality of steam without fossil superheating. 12 refs, 25 figs, 2 tabs

Study on the conversion of H2 and CO from the helium carrier gas of high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Liao Cuiping; Zheng Zhenhong; Shi Fuen

1995-01-01

The conversions of hydrogen and carbon monoxide into water vapor and carbon dioxide on CuO-ZnO-Al 2 O 3 catalyst are studied. The effects of different temperature, system atmospheric pressure, impurity gas concentration, flow and dew point on properties of cupric oxide bed are investigated. The conversion characteristics curves of H 2 and CO are given. Experimental data of conversion capacity, action period and conversion efficiency of CuO-ZnO-Al 2 O 3 are obtained and the optimal parameters are determined. The results show that the concentration of H 2 and CO of the effluent gas after purification can reach below 2 x 10 -6 , respectively. So it can meet the demands of high temperature gas-cooled reactor and also provide optimal design parameters and reliable data for conversion of H 2 and CO on CuO-ZnO-Al 2 O 3 catalyst

Initial Studies of Low Temperature Ablation in a Helium Hypersonic Wind Tunnel. Draft

Energy Technology Data Exchange (ETDEWEB)

Kohlman, D. L.; Elias, L.; Orlik-Ruckemann, K.

1969-06-15

A study of the feasibility of investigating the effects of ablation in a helium hypersonic wind tunnel was performed. Exploratory experiments were carried out at Mach 16.4 and at 600 psi stagnation pressure using (a) metal models at room temperature, (b) models with copper inserts, cooled to -140 deg C, and (c) models with carbon dioxide inserts. All models were flat plates at zero incidence, with a sharp leading edge in front of the insert. Surface temperature, surface recession rates and pitot pressure profiles were determined at several longitudinal stations. Suitable model fabrication and experimental techniques have been developed. A simple theoretical method of predicting recession rates and surface temperatures has been proposed. It has been demonstrated that the ablation of carbon dioxide into an unheated Mach 16.4 helium flow at 600 psi stagnation pressure is significant enough to result in measurable flat plate recession rates and measurable changes in pitot pressure profiles. In addition, it has been shown that it is possible to distinguish between the effects on pitot pressure of reduction in surface temperature and of mass addition through sublimation of carbon dioxide. It was also found that the first order theoretical analysis predicts proper trends and correct approximate magnitude of sublimation rates.

Mechanical Property and Its Comparison of Superalloys for High Temperature Gas Cooled Reactor

International Nuclear Information System (INIS)

Kim, Woo Gon; Kim, D. W.; Ryu, W. S.; Han, C. H.; Yoon, J. H.; Chang, J.

2005-01-01

Since structural materials for high temperature gas cooled reactor are used during long period in nuclear environment up to 1000 .deg. C, it is important to have good properties at elevated temperature such as mechanical properties (tensile, creep, fatigue, creep-fatigue), microstructural stability, interaction between metal and gas, friction and wear, hydrogen and tritium permeation, irradiation behavior, corrosion by impurity in He. Thus, in order to select excellent materials for the high temperature gas cooled reactor, it is necessary to understand the material properties and to gather the data for them. In this report, the items related to material properties which are needed for designing the high temperature gas cooled reactor were presented. Mechanical properties; tensile, creep, and fatigue etc. were investigated for Haynes 230, Hastelloy-X, In 617 and Alloy 800H, which can be used as the major structural components, such as intermediate heat exchanger (IHX), hot duct and piping and internals. Effect of He and irradiation on these structural materials was investigated. Also, mechanical properties; physical properties, tensile properties, creep and creep crack growth rate were compared for them, respectively. These results of this report can be used as important data to select superior materials for high temperature gas reactor

Modeling of helium effects in metals: High temperature embrittlement

International Nuclear Information System (INIS)

Trinkaus, H.

1985-01-01

The effects of helium on swelling, creep rupture and fatigue properties of fusion reactor materials subjected to (n,α)-reactions and/or direct α-injection, are controlled by bubble formation. The understanding of such effects requires therefore the modeling of (1) diffusional reactions of He atoms with other defects; (2) nucleation and growth of He bubbles; (3) transformation of such bubbles into cavities under continuous He generation and irradiation or creep stress. The present paper is focussed on the modeling of the (coupled) high temperature bubble nucleation and growth processes within and on grain boundaries. Two limiting cases are considered: di-atomic nucleation described by the simplest possible sets of rate equations, and multi-atomic nucleation described by classical nucleation theory. Scaling laws are derived which characterize the dependence of the bubble densities upon time (He-dose), He generation rate and temperature. Comparison with experimental data of AISI 316 SS α-implanted at temperatures around 1000 K indicates bubble nucleation of the multi-atomic type. The nucleation and growth models are applied to creep tests performed during α-implantation suggesting that in these cases gas driven bubble growth is the life time controlling mechanism. The narrow (creep stress/He generation rate) range of this mechanism in a mechanism map constructed from these tests indicates that in many reactor situations the time to rupture is probably controlled by stress driven cavity growth rather than by gas driven bubble growth. (orig.)

COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

Directory of Open Access Journals (Sweden)

Phani Kumar Domalapally

2015-04-01

Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

Magnetic properties of high temperature superconductors and their interaction with high energy permanent magnets

International Nuclear Information System (INIS)

Agarwala, A.K.

1990-01-01

Magnetic properties of sintered samples of YBCO ceramic superconductors at various temperatures were measured using a vibrating sample magnetometer (VSM). Also, measurements of forces experienced by a well characterized rare earth-transition metal (RE-TM) permanent magnet (PM) interacting with the superconducting YBCO sample cooled in liquid nitrogen, were performed. Based upon the observed hysteretic magnetization properties of these high temperature superconductors (HTS), the HTS-PM interaction force at liquid nitrogen temperature was calculated from first principle, and finally correlated to the force measurement results. With this analysis, magnetic forces between the same HTS and PM system including the levitation as well as suspension effects at liquid-helium temperature are predicted

Program for aerodynamic performance tests of helium gas compressor model of the gas turbine high temperature reactor (GTHTR300)

International Nuclear Information System (INIS)

Takada, Shoji; Takizuka, Takakazu; Kunimoto, Kazuhiko; Yan, Xing; Itaka, Hidehiko; Mori, Eiji

2003-01-01

Research and development program for helium gas compressor aerodynamics was planned for the power conversion system of the Gas Turbine High Temperature Reactor (GTHTR300). The axial compressor with polytropic efficiency of 90% and surge margin more than 30% was designed with 3-dimensional aerodynamic design. Performance and surge margin of the helium gas compressor tends to be lower due to the higher boss ratio which makes the tip clearance wide relative to the blade height, as well as due to a larger number of stages. The compressor was designed on the basis of methods and data for the aerodynamic design of industrial open-cycle gas-turbine. To validate the design of the helium gas compressor of the GTHTR300, aerodynamic performance tests were planned, and a 1/3-scale, 4-stage compressor model was designed. In the tests, the performance data of the helium gas compressor model will be acquired by using helium gas as a working fluid. The maximum design pressure at the model inlet is 0.88 MPa, which allows the Reynolds number to be sufficiently high. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

Numerical prediction on turbulent heat transfer of a spacer ribbed fuel rod for high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Takase, Kazuyuki

1994-11-01

The turbulent heat transfer of a fuel rod with three-dimensional trapezoidal spacer ribs for high temperature gas-cooled reactors was analyzed numerically using the k-ε turbulence model, and investigated experimentally using a simulated fuel rod under the helium gas condition of a maximum outlet temperature of 1000degC and pressure of 4MPa. From the experimental results, it found that the turbulent heat transfer coefficients of the fuel rod were 18 to 80% higher than those of a concentric smooth annulus at a region of Reynolds number exceeding 2000. On the other hand, the predicted average Nusselt number of the fuel rod agreed well with the heat transfer correlation obtained from the experimental data within a relative error of 10% with Reynolds number of more than 5000. It was verified that the numerical analysis results had sufficient accuracy. Furthermore, the numerical prediction could clarify quantitatively the effects of the heat transfer augmentation by the spacer rib and the axial velocity increase due to a reduction in the annular channel cross-section. (author)

Safety analysis on tokamak helium cooling slab fuel fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Wei Renjie; Jian Hongbing

1992-01-01

The thermal analyses for steady state, depressurization and total loss of flow in the tokamak helium cooling slab fuel element fusion-fission hybrid reactor are presented. The design parameters, computed results of HYBRID program and safety evaluation for conception design are given. After all, it gives some recommendations for developing the design

Study on Off-Design Steady State Performances of Helium Gas Turbo-compressor for HTGR-GT

International Nuclear Information System (INIS)

Qisen Ren; Xiaoyong Yang; Zhiyong Huang; Jie Wang

2006-01-01

The high temperature gas-cooled reactor (HTGR) coupled with direct gas turbine cycle is a promising concept in the future of nuclear power development. Both helium gas turbine and compressor are key components in the cycle. Under normal conditions, the mode of power adjustment is to control total helium mass in the primary loop using gas storage vessels. Meanwhile, thermal power of reactor core is regulated. This article analyzes off-design performances of helium gas turbine and compressors for high temperature gas-cooled reactor with gas turbine cycle (HTGR-GT) at steady state level of electric power adjustment. Moreover, performances of the cycle were simply discussed. Results show that the expansion ratio of turbine decreases as electric power reduces but the compression ratios of compressors increase, efficiencies of both turbine and compressors decrease to some extent. Thermal power does not vary consistently with electric power, the difference between these two powers increases as electric power reduces. As a result of much thermal energy dissipated in the temperature modulator set at core inlet, thermal efficiency of the cycle has a widely reduction under partial load conditions. (authors)

Desorption of tritium and helium from high dose neutron irradiated beryllium

Science.gov (United States)

Kupriyanov, I. B.; Nikolaev, G. N.; Vlasov, V. V.; Kovalev, A. M.; Chakin, V. P.

2007-08-01

The effect of high dose neutron irradiation on tritium and helium desorption in beryllium is described. Beryllium samples were irradiated in the SM and BOR-60 reactors to a neutron fluences ( E > 0.1 MeV) of (5-16) × 10 22 cm -2 at 70-100 °C and 380-420 °C. A mass-spectrometry technique was used in out of pile tritium release experiments during stepped annealing in the 250-1300 °C temperature range. The total amount of helium accumulated in irradiated beryllium samples varied from 6000 to 7200 appm. The first signs of tritium and helium release were detected at temperature of 312-445 °C and 500-740 °C, respectively. It is shown that most tritium (˜82%) from sample irradiated at 70-100 °C releases in temperature range of 312-700 °C before the beginning of helium release (740 °C). In the case of beryllium sample irradiated at 380-420 °C, tritium release starts at a higher temperature ( Ts > Tann = 445 °C) and most of the tritium (˜99.8%) is released concurrently with helium which could be considered as evidence of co-existence of partial amounts of tritium and helium in common bubbles. Both the Be samples differ little in the upper temperatures of gas release: 745 and 775 °C for tritium; 1140 and 1160 °C for helium. Swelling of beryllium starts to play a key role in accelerating tritium release at Tann > 600 °C and in helium release - at Tann > 750 °C.

Structural instabilities of high temperature alloys and their use in advanced high temperature gas cooled reactors

International Nuclear Information System (INIS)

Schuster, H.; Ennis, P.J.; Nickel, H.; Czyrska-Filemonowicz, A.

1989-01-01

High-temperature, iron-nickel and nickel based alloys are the candidate heat exchanger materials for advanced high temperature gas-cooled reactors supplying process heat for coal gasification, where operation temperatures can reach 850-950 deg. C and service lives of more than 100,000 h are necessary. In the present paper, typical examples of structural changes which occur in two representative alloys (Alloy 800 H, Fe-32Ni-20Cr and Alloy 617, Ni-22Cr-12Co-9Mo-1Al) during high temperature exposure will be given and the effects on the creep rupture properties discussed. At service temperatures, precipitation of carbides occurs which has a significant effect on the creep behaviour, especially in the early stages of creep when the precipitate particles are very fine. During coarsening of the carbides, carbides at grain boundaries restrict grain boundary sliding which retards the development of creep damage. In the service environments, enhanced carbide precipitation may occur due to the ingress of carbon from the environment (carburization). Although the creep rate is not adversely affected, the ductility of the carburized material at low and intermediate temperatures is very low. During simulated service exposures, the formation of surface corrosion scales, the precipitation of carbides and the formation of internal oxides below the surface leads to depletion of the matrix in the alloying elements involved in the corrosion processes. In thin-walled tubes the depletion of Cr due to Cr 2 O 3 formation on the surface can lead to a loss of creep strength. An additional depletion effect resulting from environmental-metal reactions is the loss of carbon (decarburization) which may occur in specific environments. The compositions of the cooling gases which decarburize the material have been determined; they are to be avoided during reactor operation

Diffusion and agglomeration of helium in stainless steel in the temperature range from RT to 600 deg. C

International Nuclear Information System (INIS)

Zhang, C.H.; Chen, K.Q.; Zhu, Z.Y.

2000-01-01

Diffusion of helium and formation of helium bubbles in stainless steel in conditions of atomic displacement in the temperature range from RT to 600 deg. C are studied theoretically using standard rate equations. The dissociative mechanism via self-interstitial/He replacement is assumed to control helium diffusion and bubble formation. The numerical analysis shows that the temperature dependence of the effective diffusion coefficient of helium, the number density and the mean radius of bubbles has two distinctly different regimes with the transition occurring around 300 deg. C. The effective diffusion coefficient of helium, the number density and the mean radius of bubbles show weak temperature dependence in the low temperature regime, while they change abruptly with temperature in the high temperature regime. The results are qualitatively in agreement with the results of our experimental study on helium diffusion and bubble formation in helium-implanted 316L stainless steel. However, the discrepancy in the absolute values of number density and mean radius of bubbles between theoretical and experimental studies indicates that helium diffusion and bubble formation may be controlled by some athermal mechanisms in the low temperature regime

State of development of high temperature gas-cooled reactors in foreign countries

International Nuclear Information System (INIS)

Sudo, Yukio

1990-01-01

Emphasis has been placed in the development of high temperature gas-cooled reactors on high thermal efficiency as power reactors and the reactor from which nuclear heat can be utilized. In U.K., as the international project 'Dragon Project', the experimental Dragon reactor for research use with 20 MWt output and exit coolant temperature 750 deg C was constructed, and operated till 1976. Coated fuel particles were developed. In West Germany, the experimental power reactor AVR with 46 MWt and 15 MWe output was operated till 1988. The prototype power reactor THTR-300 with 300 MWe output and 750 deg C exit temperature is in commercial operation. In USA, the experimental power reactor Peach Bottom reactor with 40 MWe output and 728 deg C exit temperature was operated till 1974. The prototype Fort Saint Vrain power reactor with 330 MWe output and 782 deg C exit temperature was operated till 1989. In USSR, the modular VGM with 200 MWh output is at the planning stage. Also in China, high temperature gas-cooled reactors are at the design stage. Switzerland has taken part in various international projects. (K.I.)

Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor

Directory of Open Access Journals (Sweden)

Yan Wang

2017-01-01

Full Text Available After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator.

Preliminary Analysis on Decay Heat Removal Capability of Helium Cooled Solid Breeder Test Blanket Module

International Nuclear Information System (INIS)

Ahn, Mu Young; Cho, Seung Yon; Kim, Duck Hoi; Lee, Eun Seok; Kim, Hyung Seok; Suh, Jae Seung; Yun, Sung Hwan; Cho, Nam Zin

2007-01-01

One of the main ITER goals is to test and validate design concepts of tritium breeding blankets relevant to DEMO or fusion power plants. Korea Helium-Cooled Solid Breeder (HCSB) Test Blanket Module (TBM) has been developed with overall objectives of achieving this goal. The TBM employs high pressure helium to cool down the First Wall (FW), Side Wall (SW) and Breeding Zone (BZ). Therefore, safety consideration is a part of the design process. Each ITER Party performing the TBM program is requested to reach a similar level of confidence in the TBM safety analysis. To meet ITER's request, Failure Mode and Effects Analysis (FMEA) studies have been performed on the TBM to identify the Postulated Initial Event (PIE). Although FMEA on the KO TBM has not been completed, in-vessel, in-box and ex-vessel Loss Of Coolant Accident (LOCA) are considered as enveloping cases of PIE in general. In this paper, accidental analyses for the three selected LOCA were performed to investigate the decay heat removal capability of the TBM. To simulate transient thermo-hydraulic behavior of the TBM for the selected scenarios, RELAP5/MOD3.2 code was used

Treatment and Disposal of the Radioactive Graphite Waste of High-Temperature Gas-Cooled Reactor Spent Fuel

International Nuclear Information System (INIS)

Li Junfeng

2016-01-01

High-temperature gas-cooled reactors (HTGRs) represent one of the Gen IV reactors in the future market, with efficient generation of energy and the supply of process heat at high temperature utilised in many industrial processes. HTGR development has been carried out within China’s National High Technology Research and Development Program. The first industrial demonstration HTGR of 200 MWe is under construction in Shandong Province China. HTGRs use ceramic-coated fuel particles that are strong and highly resistant to irradiation. Graphite is used as moderator and helium is used as coolant. The fuel particles and the graphite block in which they are imbedded can withstand very high temperature (up to ~1600℃). Graphite waste presents as the fuel element components of HTGR with up to 95% of the whole element beside the graphite blocks in the core. For example, a 200 MWe reactor could discharge about 90,000 fuel elements with 17 tonnes irradiated graphite included each year. The core of the HTGR in China consists of a pebble bed with spherical fuel elements. The UO 2 fuel kernel particles (0.5mm diameter) (triple-coated isotropic fuel particles) are coated by several layers including inner buffer layer with less dense pyrocarbon, dense pyro-carbon, SiC layer and outer layer of dense pyro-carbon, which can prevent the leaking of fission products (Fig. 1). Spherical fuel elements (60mm diameter) consist of a 50mm diameter inner zone and 5mm thick shell of fuel free zone [3]. The inner zone contains about 8300 triple-coated isotropic fuel particles of 0.92mm in diameter dispersed in the graphite matrix

Cryodeposition of nitrogen gas on a surface cooled by helium II

International Nuclear Information System (INIS)

Dhuley, R. C.; Bosque, E. S.; Van Sciver, S. W.

2014-01-01

Catastrophic loss of beam tube vacuum in a superconducting particle accelerator can be simulated by sudden venting of a long high vacuum channel cooled on its outer surface by He II. The rapid rush of atmospheric air in such an event shows an interesting propagation effect, which is much slower than the shock wave that occurs with vacuum loss at ambient conditions. This is due to flash frosting/deposition of air on the cold walls of the channel. Hence to characterize the propagation as well as the associated heat transfer, it is first necessary to understand the deposition process. Here we attempt to model the growth of nitrogen frost layer on a cold plate in order to estimate its thickness with time. The deposition process can be divided into two regimes- free molecular and continuum. It is shown that in free molecular regime, the frost growth can be modeled reasonably well using cryopump theory and general heat transfer relations. The continuum regime is more complex to model, given the higher rate of gas incident on cryosurface causing a large heat load on helium bath and changing cryosurface temperature. Results from the continuum regime are discussed in the context of recent experiments performed in our laboratory

Cryodeposition of nitrogen gas on a surface cooled by helium II

Energy Technology Data Exchange (ETDEWEB)

Dhuley, R. C.; Bosque, E. S.; Van Sciver, S. W. [Cryogenics Group, National High Magnetic Field Laboratory, Tallahassee, FL 32310 USA and Mechanical Engineering Department, FAMU-FSU College of Engineering, Tallahassee, FL 32310 (United States)

2014-01-29

Catastrophic loss of beam tube vacuum in a superconducting particle accelerator can be simulated by sudden venting of a long high vacuum channel cooled on its outer surface by He II. The rapid rush of atmospheric air in such an event shows an interesting propagation effect, which is much slower than the shock wave that occurs with vacuum loss at ambient conditions. This is due to flash frosting/deposition of air on the cold walls of the channel. Hence to characterize the propagation as well as the associated heat transfer, it is first necessary to understand the deposition process. Here we attempt to model the growth of nitrogen frost layer on a cold plate in order to estimate its thickness with time. The deposition process can be divided into two regimes- free molecular and continuum. It is shown that in free molecular regime, the frost growth can be modeled reasonably well using cryopump theory and general heat transfer relations. The continuum regime is more complex to model, given the higher rate of gas incident on cryosurface causing a large heat load on helium bath and changing cryosurface temperature. Results from the continuum regime are discussed in the context of recent experiments performed in our laboratory.

Technology development for the modular high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Homan, F.J.; Turner, R.F.

1989-01-01

In the USA the Modular High-Temperature Gas-Cooled Reactor is in an advanced stage of design. The related HTGR program areas, the approaches to these programs along with sample results and a description of how these data are used are highlighted in the paper. (author). Figs and tabs

Charged condensate and helium dwarf stars

Energy Technology Data Exchange (ETDEWEB)

Gabadadze, Gregory; Rosen, Rachel A, E-mail: gg32@nyu.edu, E-mail: rar339@nyu.edu [Center for Cosmology and Particle Physics, Department of Physics, New York University, New York, NY 10003 (United States)

2008-10-15

White dwarf stars composed of carbon, oxygen and heavier elements are expected to crystallize as they cool down below certain temperatures. Yet, simple arguments suggest that the helium white dwarf cores may not solidify, mostly because of zero-point oscillations of the helium ions that would dissolve the crystalline structure. We argue that the interior of the helium dwarfs may instead form a macroscopic quantum state in which the charged helium-4 nuclei are in a Bose-Einstein condensate, while the relativistic electrons form a neutralizing degenerate Fermi liquid. We discuss the electric charge screening, and the spectrum of this substance, showing that the bosonic long-wavelength fluctuations exhibit a mass gap. Hence, there is a suppression at low temperatures of the boson contribution to the specific heat-the latter being dominated by the specific heat of the electrons near the Fermi surface. This state of matter may have observational signatures.

Effects of exposure to high-temperature helium containing oxygen on the mechanical properties of molybdenum and TZM-Mo alloy at room temperature

International Nuclear Information System (INIS)

Noda, T.; Okada, M.; Watanabe, R.

1980-01-01

The effects of exposure to helium containing oxygen of 0.1-115 vpm at 1000 0 C on the mechanical properties of molybdenum and TZM-Mo alloy at room temperature were studied. The stress-relieved molybdenum specimen which was not recrystallized at test temperature showed the ductility after exposure to helium containing oxygen. The recrystallized molybdenum and TZM lost ductility after exposure to helium containing oxygen of 0.1-13 vpm in a few hours. The embrittlement of molybdenum was considered to be due to the grain boundary weakening. Molybdenum to which carbon was added seemed to hinder the grain boundary weakening by the oxygen contamination. Both stress-relieved and recrystallized TZM specimens picked up oxygen linearly with time of exposure to helium. The increase in oxygen content of TZM, which was considered to be caused by the internal oxidation of titanium and zirconium, results in the embrittlement of TZM. (orig.)

Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

International Nuclear Information System (INIS)

Shropshire, D.E.; Herring, J.S.

2004-01-01

The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

Rotary magnetic refrigerator for superfluid helium production

International Nuclear Information System (INIS)

Hakuraku, Y.; Ogata, H.

1986-01-01

A new rotary-magnetic refrigerator designed to obtain superfluid helium temperatures by executing a magnetic Carnot cycle is developed. A rotor containing 12 magnetic refrigerants (gadolinium-gallium-garnet) is immersed in liquid helium at 4.2 K and rotated at constant speed in a steady magnetic field distribution. Performance tests demonstrate that the new rotary refrigerator is capable of obtaining a temperature of 1.48 K. The maximum useful cooling power obtained at 1.8 K is 1.81 W which corresponds to a refrigeration efficiency of 34%

Development of analytical code `ACCORD` for incore and plant dynamics of High Temperature Gas-cooled Reactor

Energy Technology Data Exchange (ETDEWEB)

Takeda, Takeshi; Tachibana, Yukio; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Itakura, Hirofumi

1996-11-01

Safety demonstration test of the High Temperature Engineering Test Reactor will be carried out to demonstrate excellent safety features of a next generation High Temperature Gas-cooled Reactor (HTGR). Analytical code for incore and plant dynamics is necessary to assess the results of the safety demonstration test and to perform a design and safety analysis of the next generation HTGR. Existing analytical code for incore and plant dynamics of the HTGR can analyze behavior of plant system for only several thousand seconds after an event occurrence. Simulator on site can analyze only behavior of specific plant system. The `ACCORD` code has been, therefore, developed to analyze the incore and plant dynamics of the HTGR. The followings are the major characteristics of this code. (1) Plant system can be analyzed for over several thousand seconds after an event occurrence by modeling the heat capacity of the core. (2) Incore and plant dynamics of any plant system can be analyzed by rearranging packages which simulate plant system components one by one. (3) Thermal hydraulics for each component can be analyzed by separating heat transfer calculation for component from fluid flow calculation for helium and pressurized water systems. The validity of the `ACCORD` code including models for nuclear calculation, heat transfer and fluid flow calculation, control system and safety protection system, was confirmed through cross checks with other available codes. (author)

Development of analytical code 'ACCORD' for incore and plant dynamics of High Temperature Gas-cooled Reactor

International Nuclear Information System (INIS)

Takeda, Takeshi; Tachibana, Yukio; Kunitomi, Kazuhiko; Itakura, Hirofumi.

1996-11-01

Safety demonstration test of the High Temperature Engineering Test Reactor will be carried out to demonstrate excellent safety features of a next generation High Temperature Gas-cooled Reactor (HTGR). Analytical code for incore and plant dynamics is necessary to assess the results of the safety demonstration test and to perform a design and safety analysis of the next generation HTGR. Existing analytical code for incore and plant dynamics of the HTGR can analyze behavior of plant system for only several thousand seconds after an event occurrence. Simulator on site can analyze only behavior of specific plant system. The 'ACCORD' code has been, therefore, developed to analyze the incore and plant dynamics of the HTGR. The followings are the major characteristics of this code. (1) Plant system can be analyzed for over several thousand seconds after an event occurrence by modeling the heat capacity of the core. (2) Incore and plant dynamics of any plant system can be analyzed by rearranging packages which simulate plant system components one by one. (3) Thermal hydraulics for each component can be analyzed by separating heat transfer calculation for component from fluid flow calculation for helium and pressurized water systems. The validity of the 'ACCORD' code including models for nuclear calculation, heat transfer and fluid flow calculation, control system and safety protection system, was confirmed through cross checks with other available codes. (author)

Helium compressor aerodynamic design considerations for MHTGR circulators

International Nuclear Information System (INIS)

McDonald, C.F.

1988-01-01

Compressor aerodynamic design considerations for both the main and shutdown cooling circulators in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) plant are addressed in this paper. A major selection topic relates to the impeller type (i.e., axial or radial flow), and the aerothermal studies leading to the selection of optimum parameters are discussed. For the conceptual designs of the main and shutdown cooling circulators, compressor blading geometries were established and helium gas flow paths defined. Both circulators are conservative by industrial standards in terms of aerodynamic and structural loading, and the blade tip speeds are particularly modest. Performance characteristics are presented, and the designs embody margin to ensure that pressure-rise growth potential can be accomodated should the circuit resistance possibly increase as the plant design advances. The axial flow impeller for the main circulator is very similar to the Fort St. Vrain (FSV) helium compressor which performs well. A significant technology base exists for the MHTGR plant circulators, and this is highlighted in the paper. (author). 15 refs, 16 figs, 12 tabs

Study on extreme high temperature of cooling water in Chinese coastal nuclear power plant

International Nuclear Information System (INIS)

Yu Fan; Jiang Ziying

2012-01-01

In order to protect aquatic life from the harmful effects of thermal discharge, the appropriate water temperature limits or the scope of the mixing zone is a key issue in the regulatory control of the environmental impact of thermal discharge. Based on the sea surface temperature in the Chinese coastal waters, the extreme value of the seawater temperature change was analyzed by using the Gumbel model. The limit of the design temperature rise of cooling water in the outfall is 9 ℃, and the limit of the temperature rise of cooling water in the edge of the mixing zone is 4 ℃. The extreme high temperature of the cooling water in Chinese coastal nuclear power plant is 37 ℃ in the Bohai Sea, Yellow Sea, and is 40 ℃ in East China Sea, South China Sea. (authors)

Study on restriction method for end-wall boundary layer thickness in axial helium gas compressor for gas turbine high temperature gas cooled reactor

International Nuclear Information System (INIS)

Takada, Shoji; Takizuka, Takakazu; Yan, Xing; Kunitomi, Kazuhiko; Inagaki, Yoshiyuki

2009-01-01

Aerodynamic performance test was carried out using a 1/3 scale, 4-stage model of the helium gas compressor to investigate an effect of end-wall over-camber to prevent decrease of axial velocity in the end-wall boundary layer. The model compressor consists of a rotor, 500 mm in diameter, which is driven by an electric motor at a rotational speed of 10800 rpm. The rotor blade span of the first stage is 34 mm. The test was carried out under the condition that the helium gas pressure of 0.88 MPa, temperature of 30degC, and mass flow rate of 12.47 kg/s at the inlet. A 3-dimensional aerodynamic code, which was verified using the test data, showed that axial velocity was lowered by using a blade which increased the inlet blade angle around the end-wall region of the casing side in comparison with that using the original design blade, because the inlet flow angle mismatched with the inlet blade angle of the rotor blade, as opposed to the prediction by a conventional air compressor design method. The overall adiabatic efficiency of the full scale 20-stage helium gas compressor was predicted 89.7% from the Reynolds number dependency of the test data by using the original design blade. (author)

Discussion on Design Transients of Pebble-bed High Temperature Gas-cooled Reactor

International Nuclear Information System (INIS)

Wang Yan; Li Fu; Zheng Yanhua

2014-01-01

In order to assure high quality for the components and their supports in the reactor coolant system, etc., some thermal-hydraulic transient conditions will be selected and researched for equipment design evaluation to satisfy the requirements ASME code, which are based on the conservative estimates of the magnitude and frequency of the temperature and pressure transients resulting from various operating conditions in the plant. In the mature design on pressurized water reactor, five conditions are considered. For the developing advanced pebble-bed high temperature gas-cooled reactor(HTGR), its design and operation has much difference with other reactors, so the transients of the pebble-bed high temperature gas-cooled reactor have distinctive characteristics. In this paper, the possible design transients of the pebble-bed HTGR will be discussed, and the frequency of design transients for equipment fatigue analysis and stress analysis due to cyclic stresses is also studied. The results will provide support for the design and construct of the pebble-bed HTGR. (author)

DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-[TEMPERATURE GAS-COOLED TEST REACTOR

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

2016-11-01

A point design for a graphite-moderated, high-temperature, gas-cooled test reactor (HTG TR) has been developed by Idaho National Laboratory (INL) as part of a United States (U.S.) Department of Energy (DOE) initiative to explore and potentially expand the existing U.S. test reactor capabilities. This paper provides a summary of the design and its main attributes. The 200 MW HTG TR is a thermal-neutron spectrum reactor composed of hexagonal prismatic fuel and graphite reflector blocks. Twelve fuel columns (96 fuel blocks total and 6.34 m active core height) are arranged in two hexagonal rings to form a relatively compact, high-power density, annular core sandwiched between inner, outer, top, and bottom graphite reflectors. The HTG-TR is designed to operate at 7 MPa with a coolant inlet/outlet temperature of 325°C/650°C, and utilizes TRISO particle fuel from the DOE AGR Program with 425 ?m uranium oxycarbide (UCO) kernels and an enrichment of 15.5 wt% 235U. The primary mission of the HTG TR is material irradiation and therefore the core has been specifically designed and optimized to provide the highest possible thermal and fast neutron fluxes. The highest thermal neutron flux (3.90E+14 n/cm2s) occurs in the outer reflector, and the maximum fast flux levels (1.17E+14 n/cm2s) are produced in the central reflector column where most of the graphite has been removed. Due to high core temperatures under accident conditions, all the irradiation test facilities have been located in the inner and outer reflectors where fast flux levels decline. The core features a large number of irradiation positions with large test volumes and long test lengths, ideal for thermal neutron irradiation of large test articles. The total available test volume is more than 1100 liters. Up to four test loop facilities can be accommodated with pressure tube boundaries to isolate test articles and test fluids (e.g., liquid metal, liquid salt, light water) from the helium primary coolant system.

High temperature gas-cooled reactor: gas turbine application study

Energy Technology Data Exchange (ETDEWEB)

1980-12-01

The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

High temperature gas-cooled reactor: gas turbine application study

International Nuclear Information System (INIS)

1980-12-01

The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project

Potentialities in electronics of new high critical temperature superconductors. Potentialites en electronique des nouveaux supraconducteurs a haute temperature critique

Energy Technology Data Exchange (ETDEWEB)

Hartemann, P [Thomson-CSF, 75 - Paris (FR)

1989-09-01

The main electronic applications of superconductors involve the signal processing, the electromagnetic wave detection and the magnetometry. Characteristics of devices based on conventional superconductors cooled by liquid helium are given and the changes induced by incorporating high-temperature superconductors are estimated. After a survey of new superconductor properties, the superconducting devices for analog or digital signal processing are reviewed. The gains predicted for high-temperature superconducting analog devices are considered in greater detail. Different sections deal with the infrared or (sub)millimeter wave detection. The most sensitive apparatuses for magnetic measurements are based on SQUIDs. Features of SQUIDs made of granular high-temperature superconducting material samples (grain boundaries behave as barriers of intrinsic junctions) are discussed.

The United States fluoride-salt-cooled high-temperature reactor program

International Nuclear Information System (INIS)

Holcomb, David E.

2013-01-01

The United States is pursuing the development of fluoride-salt-cooled high-temperature reactors (FHRs) through the Department of Energy's Office of Nuclear Energy (DOE-NE). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. FHRs, in principle, have the potential to economically generate large amounts of electricity while maintaining full passive safety. FHRs, however, remain a longer-term power production option. A principal development focus is, thus, on shortening, to the extent possible, the overall development time by focusing initial efforts on the longest lead-time issues. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid-metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High-temperature gas-cooled reactors provide experience with coated-particle fuel and graphite components. Light-water reactors show the potential of transparent, high-heat-capacity coolants with low chemical reactivity. The FHR development efforts include both reactor concept and technology developments and are being broadly pursued. Oak Ridge National Laboratory (ORNL) provides technical leadership to the effort and is performing concept development on both a large base-load-type FHR as well as a small modular reactor (SMR) in addition to performing a broad scope of technology developments. Idaho National Laboratory (INL) is providing coated-particle fuel irradiation testing as well as developing high-temperature steam generator technology. The Massachusetts Institute of Technology (MIT

Design and study of Engineering Test Facility - Helium Circulator

International Nuclear Information System (INIS)

Jiang Huijing; Ye Ping; Zhao Gang; Geng Yinan; Wang Jie

2015-01-01

Helium circulator is one of the key equipment of High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). In order to simulate most normal and accident operating conditions of helium circulator in HTR-PM, a full scale, rated flow rate and power, engineering test loop, which was called Engineering Test Facility - Helium Circulator (ETF-HC), was designed and established. Two prototypes of helium circulator, which was supported by Active Magnetic Bearing (AMB) or sealed by dry gas seals, would be tested on ETF-HC. Therefore, special interchangeable design was under consideration. ETF-HC was constructed compactly, which consisted of eleven sub-systems. In order to reduce the flow resistance of the circuit, special ducts, elbows, valves and flowmeters were selected. Two stages of heat exchange loops were designed and a helium - high pressure pure water heat exchanger was applied to ensure water wouldn't be vaporized while simulating accident conditions. Commissioning tests were carried out and operation results showed that ETF-HC meets the requirement of helium circulator operation. On this test facility, different kinds of experiments were supposed to be held, including mechanical and aerodynamic performance tests, durability tests and so on. These tests would provide the features and performance of helium circulator and verify its feasibility, availability and reliability. (author)

Baseline Concept Description of a Small Modular High Temperature Reactor

Energy Technology Data Exchange (ETDEWEB)

Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-10-01

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

Baseline Concept Description of a Small Modular High Temperature Reactor

Energy Technology Data Exchange (ETDEWEB)

Hans Gougar

2014-05-01

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

New deployment of high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Sekimoto, Hiroshi; Tsuchie, Yasuo; Kunitomi, Kazuhiko; Shiozawa, Shusaku; Konuki, Kaoru; Inagaki, Yoshiyuki; Hayakawa, Hitoshi

2002-01-01

The high temperature gas-cooled reactor (HTGR) is now under a condition difficult to know it well, because of considering not only power generation, but also diverse applications of its nuclear heat, of having extremely different safe principle from that of conventional reactors, of having two types of pebble-bed and block which are extremely different types, of promoting its construction plan in South Africa, of including its application to disposition of Russian surplus weapons plutonium of less reporting HTTR in Japan in spite of its full operation, and so on. However, HTGR is expected for an extremely important nuclear reactor aiming at the next coming one of LWR. HTGR which is late started and developed under complete private leading, is strongly conscious at environmental problem since its beginning. Before 30 years when large scale HTGR was expected to operate, it advertised a merit to reduce wasted heat because of its high temperature. As ratio occupied by electricity expands among application of energies, ratio occupied by the other energies are larger. When considering applications except electric power, high temperature thermal energy from HTGR can be thought wider applications than that from LWR and so on. (G.K.)

Cooling System Design Options for a Fusion Reactor

Science.gov (United States)

Natalizio, Antonio; Collén, Jan; Vieider, Gottfried

1997-06-01

The objective of a fusion power reactor is to produce electricity safely and reliably. Accordingly, the design, objective of the heat transport system is to optimize power production, safety, and reliability. Such an optimization process, however, is constrained by many factors, including, among others: public safety, worker safety, steam cycle efficiency, reliability, and cost. As these factors impose conflicting requirements, there is a need to find an optimum design solution, i.e., one that satisfies all requirements, but not necessarily each requirement optimally. The SEAFP reactor study developed helium-cooled and water-cooled models for assessment purposes. Among other things, the current study demonstrates that neither model offers an optimum solution. Helium cooling offers a high steam cycle efficiency but poor reliability for the cooling of high heat flux components (divertor and first wall). Alternatively, water cooling offers a low steam cycle efficiency, but reasonable reliability for the cooling of such components. It is concluded that an optimum solution includes helium cooling of low heat flux components and water cooling of high heat flux components. Relative to the SEAFP helium model, this hybrid system enhances safety and reliability, while retaining the high steam cycle efficiency of that model.

Construction of helium engineering demonstration loop (HENDEL M+A) for VHTR

International Nuclear Information System (INIS)

Shimomura, Saneaki; Tanaka, Toshiyuki; Nakano, Tadasuke

1983-01-01

The mother and adapter sections of the large structural component demonstration test loop, alias Helium Engineering Demonstration Loop, for the multipurpose, high temperature gas-cooled experimental reactor were completed in March, 1982. This facility was constructed by Fuji Electric Co., Ltd. and Kawasaki Heavy Industries Ltd. as the main contractors, and by the cooperation with Mitsubishi Heavy Industries Ltd. and Ishikawajima Harima Heavy Industries Co., Ltd. The HENDEL M+A is the testing facility of the largest scale in the world, which can handle 1000 deg C, 40 kgf/cm 2 G helium at a half flow rate of one core cooling loop of the experimental reactor. With the HENDEL M+A, the demonstration tests of fuel assembly stacks, in-core structures, large flow rate and high temperature equipment are planned. The HENDEL M+A comprises two mother loops, an adapter loop, and common auxiliary systems fon measurement and control (In), refining (Mp), makeup (Mu) and cooling water (Uc). The construction and function of such main equipment as a heater, circulators and internally insulated piping are described. The progress of the construction and the main experience during the construction, the process of operation and the performance are reported. (Kako, I.)

Low temperature high frequency coaxial pulse tube for space application

Energy Technology Data Exchange (ETDEWEB)

Charrier, Aurelia; Charles, Ivan; Rousset, Bernard; Duval, Jean-Marc [SBT, UMR-E CEA / UJF-Grenoble 1, INAC, 17, rue des Martyrs, Grenoble, F-38054 (France); Daniel, Christophe [CNES, 18, avenue Edouard Belin, Toulouse, F-31401 (France)

2014-01-29

The 4K stage is a critical step for space missions. The Hershel mission is using a helium bath, which is consumed day by day (after depletion, the space mission is over) while the Plank mission is equipped with one He4 Joule-Thomson cooler. Cryogenic chain without helium bath is a challenge for space missions and 4.2K Pulse-Tube working at high frequency (around 30Hz) is one option to take it up. A low temperature Pulse-Tube would be suitable for the ESA space mission EChO (Exoplanet Characterisation Observatory, expected launch in 2022), which requires around 30mW cooling power at 6K; and for the ESA space mission ATHENA (Advanced Telescope for High ENergy Astrophysics), to pre-cool the sub-kelvin cooler (few hundreds of mW at 15K). The test bench described in this paper combines a Gifford-McMahon with a coaxial Pulse-Tube. A thermal link is joining the intercept of the Pulse-Tube and the second stage of the Gifford-McMahon. This intercept is a separator between the hot and the cold regenerators of the Pulse-Tube. The work has been focused on the cold part of this cold finger. Coupled with an active phase shifter, this Pulse-Tube has been tested and optimized and temperatures as low as 6K have been obtained at 30Hz with an intercept temperature at 20K.

Relative thermoluminescent efficiencies proton/gamma and helium/gamma of high temperature peaks in TLD-100 dosemeters

International Nuclear Information System (INIS)

Flores M, E.; Avila, O.; Rodriguez V, M.; Massillon, J.L.G.; Buenfil A, E.; Ruiz T, C.; Brandan, M.E.; Gamboa De Buen, I.

2007-01-01

This work presents measures of relative thermoluminescent efficiency of those high temperature peaks of TLD-100 dosemeters exposed to protons of 1.5 MeV and to helium nuclei of 3 and 7.5 MeV. A rigorous reading and of deconvolution protocol was used for the calculation of the TL efficiencies. Additionally an Excel program that facilitated the deconvolution adjustment process of the glow curves was elaborated. (Author)

High resolution spectroscopy of six new extreme helium stars

Science.gov (United States)

Heber, U.; Jones, G.; Drilling, J. S.

1986-01-01

High resolution spectra of six newly discovered extreme helium stars are presented. LSS 5121 is shown to be a spectroscopical twin of the hot extreme helium star HD 160641. A preliminary LTE analysis of LSS 3184 yielded an effective temperature of 22,000 K and a surface gravity of log g = 3.2. Four stars form a new subgroup, classified by sharp-lined He I spectra and pronounced O II spectra, and it is conjectured that these lie close to the Eddington limit. The whole group of extreme helium stars apparently is inhomogeneous with respect to luminosity to mass ratio and chemical composition.

Liquid helium plant in Dubna

International Nuclear Information System (INIS)

Agapov, N.N.; Baldin, A.M.; Kovalenko, A.D.

1995-01-01

The liquid-helium cooling capacity installed at the Laboratory of High Energies is about 5 kw at a 4.5 K temperature level. It is provided with four industrial helium liquefiers of 1.6 kw/4.5 K each. They have been made by the Russian enterprise NPO GELYMASH and upgraded by the specialists of the Laboratory. The first one was put into operation in 1980, the two others in 1991, and the last one is under commissioning. The development of the LHE cryoplant was concerned with the construction of the new superconducting accelerator Nuclotron aimed to accelerate nuclei and heavy ions up to energies of 6 GeV/u. The first test run at the Nuclotron was carried out in March 1993, and the total running time has been about 2000 hours up to now. Since 1992 the cryoplant has been intensively used by the users outside the Laboratory. More than a million liters of liquid helium was provided in 1993 for such users. The reliability of the cryoplant system was as high as 98 percent for 4500 hours of operation in 1993-1994. 7 refs., 4 figs., 1 tab

Application of assembly module to high-temperature gas-cooled reactor full-scope simulation system

International Nuclear Information System (INIS)

Li Sifeng; Li Fu; Ma Yuanle; Shi Lei

2007-01-01

According to the circumstances that exist in the reactor full-scope simulators development as long development cycle, very difficult upgrade and narrow range of applicability, a kind of new model was developed based on assembly module which root in Linux kernel and successfully applied to the design of high-temperature gas-cooled reactor full-scope simulator system. The simulation results are coincident with the experimental ones, and it indicates that the new model based on assembly module is feasible to design of high-temperature gas cooled reactor simulation system. (authors)

Utility industry evaluation of the Modular High-Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Burstein, S.; Bitel, J.S.; Tramm, T.R.; High, M.D.; Neils, G.H.; Tomonto, J.R.; Weinberg, C.J.

1990-02-01

A team of utility industry representatives evaluated the Modular High Temperature Gas-Cooled Reactor plant design, a current design created by an industrial team led by General Atomics under Department of Energy sponsorship and with support provided by utilities through Gas-Cooled Reactor Associates. The utility industry team concluded that the plant design should be considered a viable application of an advanced nuclear concept and deserves continuing development. Specific comments and recommendations are provided as a contribution toward improving a very promising plant design. 2 refs

Advanced Gas Cooled Reactor Materials Program. Reducing helium impurity depletion in HTGR materials testing

International Nuclear Information System (INIS)

Baldwin, D.H.

1984-08-01

Moisture depletion in HTGR materials testing rigs has been empirically studied in the GE High Temperature Reactor Materials Testing Laboratory (HTRMTL). Tests have shown that increased helium flow rates and reduction in reactive (oxidizable) surface area are effective means of reducing depletion. Further, a portion of the depletion has been shown to be due to the presence of free C released by the dissociation of CH 4 . This depletion component can be reduced by reducing the helium residence time (increasing the helium flow rate) or by reducing the CH 4 concentration in the test gas. Equipment modifications to reduce depletion have been developed, tested, and in most cases implemented in the HTRMTL to date. These include increasing the Helium Loop No. 1 pumping capacity, conversion of metallic retorts and radiation shields to alumina, isolation of thermocouple probes from the test gas by alumina thermowells, and substitution of non-reactive Mo-TZM for reactive metallic structural components

Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

NARCIS (Netherlands)

van Rooijen, W.F.G.

2006-01-01

This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

Modeling and performance of the MHTGR [Modular High-Temperature Gas-Cooled Reactor] reactor cavity cooling system

International Nuclear Information System (INIS)

Conklin, J.C.

1990-04-01

The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab

Economical evaluation on gas turbine high temperature reactor 300 (GTHTR300)

International Nuclear Information System (INIS)

Takei, Masanobu; Kosugiyama, Shinichi; Mouri, Tomoaki; Katanishi, Shoji; Kunitomi, Kazuhiko

2006-01-01

Japan Atomic Energy Research Institute (JAERI) has been developing a graphite moderate and helium cooled High Temperature Gas-cooled Reactor (HTGR) with gas turbine, the GTHTR300 based on experience gained in development and operations of the High Temperature Engineering Test Reactor (HTTR) in JAERI. The GTHTR300 is a simplified and economical power plant with a high level of safety characteristics and a high plant efficiency of approximately 46%. Cost evaluation for plant construction and power generation is studied in order to clarify the economical feasibility of the GTHTR300. The construction cost is estimated to be about 200 thousands Yen/kWe. The power generation cost is estimated to be about 3.8 Yen/kWh by the conditions of 90% load factor and 3% discount rate. The economical feasibility of the GTHTR300 is certified. The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

Design of high temperature Engineering Test Reactor (HTTR)

International Nuclear Information System (INIS)

Saito, Shinzo; Tanaka, Toshiyuki; Sudo, Yukio

1994-09-01

Construction of High Temperature Engineering Test Reactor (HTTR) is now underway to establish and upgrade basic technologies for HTGRs and to conduct innovative basic research at high temperatures. The HTTR is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 850degC for rated operation and 950degC for high temperature test operation. It is planned to conduct various irradiation tests for fuels and materials, safety demonstration tests and nuclear heat application tests. JAERI received construction permit of HTTR reactor facility in February 1990 after 22 months of safety review. This report summarizes evaluation of nuclear and thermal-hydraulic characteristics, design outline of major systems and components, and also includes relating R and D result and safety evaluation. Criteria for judgment, selection of postulated events, major analytical conditions for anticipated operational occurrences and accidents, computer codes used in safety analysis and evaluation of each event are presented in the safety evaluation. (author)

Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

International Nuclear Information System (INIS)

Cui, Shijie; Zhang, Dalin; Cheng, Jie; Tian, Wenxi; Su, G.H.

2017-01-01

As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

Energy Technology Data Exchange (ETDEWEB)

Cui, Shijie; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Cheng, Jie; Tian, Wenxi; Su, G.H.

2017-01-15

As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

Assessments of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Zhang Zuoyi; Dong Yujie; Scherer, Winfried

2005-01-01

Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemented in the dynamic simulator for nuclear power plants (DSNP) simulation system. Comparisons of the DSNP simulations incorporating these models with experiments and with calculations using the time-dependent neutronics and temperature dynamics code were made to validate the simulation. The analysis of the primary circuit showed that the maximum water concentration increase in the reactor core was 3 s). The water vaporization in the steam generator and characteristics of water transport from the steam generator to the reactor core would reduce the rate of water ingress into the reactor core. The analysis of a full cavitation of the feedwater pump showed that if the secondary circuit could be depressurized, the feedwater pump would be stopped by the full cavitation. This limits the water transported from the deaerator to the steam generator. A comprehensive simulation of the HTR-module power plant showed that the water inventory in the primary circuit was limited to ∼3000 kg. The nuclear reactivity increase caused by the water ingress would lead to a fast power excursion, which would be inherently counterbalanced by negative feedback effects. The integrity of the fuel elements, because the safety-relevant temperature limit of 1600 deg. C is not reached in any case, is not challenged

CFD Analysis on the Passive Heat Removal by Helium and Air in the Canister of Spent Fuel Dry Storage System

Energy Technology Data Exchange (ETDEWEB)

Shin, Do Young; Jeong, Ui Ju; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

2016-05-15

In the current commercial design, the canister of the dry storage system is mainly backfilled with helium gas. Helium gas shows very conductive behavior due to high thermal conductivity and small density change with temperature. However, other gases such as air, argon, or nitrogen are expected to show effective convective behavior. Thus these are also considered as candidates for the backfill gas to provide effective coolability. In this study, to compare the dominant cooling mechanism and effectiveness of cooling between helium gas and air, a computational fluid dynamics (CFD) analysis for the canister of spent fuel dry storage system with backfill gas of helium and air is carried out. In this study, CFD simulations for the helium and air backfilled gas for dry storage system canister were carried out using ANSYS FLUENT code. For the comparison work, two backfilled fluids were modeled with same initial and boundary conditions. The observed major difference can be summarized as follows. - The simulation results showed the difference in dominant heat removal mechanism. Conduction for helium, and convection for air considering Reynolds number distribution. - The temperature gradient inside the fuel assembly showed that in case of air, more effective heat mixing occurred compared to helium.

A dynamic model of an innovative high-temperature solar heating and cooling system

Directory of Open Access Journals (Sweden)

Buonomano Annamaria

2016-01-01

Full Text Available In this paper a new simulation model of a novel solar heating and cooling system based on innovative high temperature flat plate evacuated solar thermal collector is presented. The system configuration includes: flat-plate evacuated solar collectors, a double-stage LiBr-H2O absorption chiller, gas-fired auxiliary heater, a closed loop cooling tower, pumps, heat exchangers, storage tanks, valves, mixers and controllers. The novelty of this study lies in the utilization of flat-plate stationary solar collectors, manufactured by TVP Solar, rather than concentrating ones (typically adopted for driving double-stage absorption chillers. Such devices show ultra-high thermal efficiencies, even at very high (about 200°C operating temperatures, thanks to the high vacuum insulation. Aim of the paper is to analyse the energy and economic feasibility of such novel technology, by including it in a prototypal solar heating and cooling system. For this purpose, the solar heating and cooling system design and performance were analysed by means of a purposely developed dynamic simulation model, implemented in TRNSYS. A suitable case study is also presented. Here, the simulated plant is conceived for the space heating and cooling and the domestic hot water production of a small building, whose energy needs are fulfilled through a real installation (settled also for experimental purposes built up close to Naples (South Italy. Simulation results show that the investigated system is able to reach high thermal efficiencies and very good energy performance. Finally, the economic analysis shows results comparable to those achieved through similar renewable energy systems.

Helium solubility and bubble growth in metals under high pressure

International Nuclear Information System (INIS)

Laakmann, J.

1985-07-01

Helium solubility and bubble growth in metals under high pressure polycrystals and single crystals of gold were heated in helium at temperatures between 475 K and 1250 K in a pressure regime of 200 to 2700 bar to measure the solubility of helium in gold. After quenching to room temperature the helium content, measured by mass spectrometry, showed the following properties: 1) A linear dependence of the He solubility on pressure. 2) Thinning of the specimen reduces the helium content by a factor 10 to 100 but does not change the linear pressure dependence. 3) The thermal release of He from thinned polycrystals and single crystals occurs mainly in a single peak at 500 K. 4) The He concentration of the thinned single crystals was lower by a factor of 10 to 50 than that of the thinned polycrystals. 5) The He solubility in single crystals can be described by an enthalpy of solution Hsub(s)sup(f) = 0.85 +- 0.7 eV and a non-configurational entropy of Ssub(s)sup(f) between 0 k and 1 k (k: Boltzmann-constant). In order to measure the pressure dependence of helium bubble growth in nickel polycrystal Ni-foils were α-implanted to a helium content of 130 appm. The evaluation of the size distribution of the helium bubbles after heat treatments shows 1) The helium content of the observable bubbles - assumed to be in equilibrium - equals the amount of helium implanted into the specimen. 2) The activation energy for the growth of helium bubbles is 1.25 +- 0.3 eV. The comparison of specimen which had been heated at low pressures up to 10 bar with others heated at 2500-2700 bar does not show an unequivocal pressure dependence for helium bubble growth. (orig./IHOE) [de

Research program of the high temperature engineering test reactor for upgrading the HTGR technology

International Nuclear Information System (INIS)

Kunitomi, Kazuhiko; Tachibana, Yukio; Takeda, Takeshi; Saikusa, Akio; Sawa, Kazuhiro

1997-07-01

The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium-cooled reactor with an outlet power of 30 MW and outlet coolant temperature of 950degC, and its first criticality will be attained at the end of 1997. In the HTTR, researches establishing and upgrading the technology basis necessary for an HTGR and innovative basic researches for a high temperature engineering will be conducted. A research program of the HTTR for upgrading the technology basis for the HTGR was determined considering realization of future generation commercial HTGRs. This paper describes a research program of the HTTR. (author)

Investigation of heating and cooling in a stand-alone high temperature PEM fuel cell system

International Nuclear Information System (INIS)

Zhang, Caizhi; Yu, Tao; Yi, Jun; Liu, Zhitao; Raj, Kamal Abdul Rasheedj; Xia, Lingchao; Tu, Zhengkai; Chan, Siew Hwa

2016-01-01

Highlights: • Heating-up and cooling-down processes of HT-PEMFC are the mainly interested topics. • Dynamic behaviours, power and energy demand of the heating and cooling was studied. • Hybrid system based on LiFeYPO_4 battery for heating and cooling is built and tested. • The concept of combining different heating sources together is recommended. - Abstract: One key issue pertaining to the cold-start of High temperature PEM fuel cell (HT-PEMFC) is the requirement of high amount of thermal energy for heating up the stack to a temperature of 120 °C or above before it can generate electricity. Furthermore, cooling down the stack to a certain temperature (e.g. 50 °C) is necessary before stopping. In this study, the dynamic behaviours, power and energy demand of a 6 kW liquid cooled HT-PEMFC stack during heating-up, operation and cooling-down were investigated experimentally. The dynamic behaviours of fuel cell under heating-up and cooling-down processes are the mainly interested topics. Then a hybridisation of HT-PEMFC with Li-ion battery to demonstrate the synergistic effect on dynamic behaviour was conducted and validated for its feasibility. At last, the concept of combining different heating sources together is analysed to reduce the heating time of the HT-PEMFC as well.

Optomechanics in a Levitated Droplet of Superfluid Helium

Science.gov (United States)

Brown, Charles; Harris, Glen; Harris, Jack

2017-04-01

A critical issue common to all optomechanical systems is dissipative coupling to the environment, which limits the system's quantum coherence. Superfluid helium's extremely low optical and mechanical dissipation, as well as its high thermal conductivity and its ability cool itself via evaporation, makes the mostly uncharted territory of superfluid optomechanics an exciting avenue for exploring quantum effects in macroscopic objects. I will describe ongoing work that aims to exploit the unique properties of superfluid helium by constructing an optomechanical system consisting of a magnetically levitated droplet of superfluid helium., The optical whispering gallery modes (WGMs) of the droplet, as well as the mechanical oscillations of its surface, should offer exceptionally low dissipation, and should couple to each other via the usual optomechanical interactions. I will present recent progress towards this goal, and also discuss the background for this work, which includes prior demonstrations of magnetic levitation of superfluid helium, high finesse WGMs in liquid drops, and the self-cooling of helium drops in vacuum.

Concept of an inherently-safe high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

2012-01-01

As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of the confinement function are identified. It is proposed not to apply the probabilistic approach for the evaluation of the radiological consequences of the accidents in the safety analysis because no inherent safety features fail for the mitigation of the consequences of the accidents. Consequently, there are no event sequences to harmful release of radioactive materials if the design extension conditions occur in the inherently-safe HTGR concept. The concept and future R and D items for the inherently-safe HTGR are described in this paper.

Gas-cooled Fast Reactor (GFR) fuel and In-Core Fuel Management

International Nuclear Information System (INIS)

Weaver, K.D.; Sterbentz, J.; Meyer, M.; Lowden, R.; Hoffman, E.; Wei, T.Y.C.

2004-01-01

The Gas-Cooled Fast Reactor (GCFR) has been chosen as one of six candidates for development as a Generation IV nuclear reactor based on: its ability to fully utilize fuel resources; minimize or reduce its own (and other systems) actinide inventory; produce high efficiency electricity; and the possibility to utilize high temperature process heat. Current design approaches include a high temperature (2 850 C) helium cooled reactor using a direct Brayton cycle, and a moderate temperature (550 C - 650 C) helium or supercritical carbon dioxide (S-CO 2 ) cooled reactor using direct or indirect Brayton cycles. These design choices have thermal efficiencies that approach 45% to 50%, and have turbomachinery sizes that are much more compact compared to steam plants. However, there are challenges associated with the GCFR, which are the focus of current research. This includes safety system design for decay heat removal, development of high temperature/high fluence fuels and materials, and development of fuel cycle strategies. The work presented here focuses on the fuel and preliminary in-core fuel management, where advanced ceramic-ceramic (cercer) dispersion fuels are the main focus, and average burnups to 266 M Wd/kg appear achievable for the reference Si C/(U,TRU)C block/plate fuel. Solid solution (pellet) fuel in composite ceramic clad (Si C/Si C) is also being considered, but remains as a backup due to cladding fabrication challenges, and high centerline temperatures in the fuel. (Author)

Transient boiling in two-phase helium natural circulation loops

Science.gov (United States)

Furci, H.; Baudouy, B.; Four, A.; Meuris, C.

2014-01-01

Two-phase helium natural circulation loops are used for cooling large superconducting magnets, as CMS for LHC. During normal operation or in the case of incidents, transients are exerted on the cooling system. Here a cooling system of this type is studied experimentally. Sudden power changes are operated on a vertical-heated-section natural convection loop, simulating a fast increase of heat deposition on magnet cooling pipes. Mass flow rate, heated section wall temperature and pressure drop variations are measured as a function of time, to assess the time behavior concerning the boiling regime according to the values of power injected on the heated section. The boiling curves and critical heat flux (CHF) values have been obtained in steady state. Temperature evolution has been observed in order to explore the operating ranges where heat transfer is deteriorated. Premature film boiling has been observed during transients on the heated section in some power ranges, even at appreciably lower values than the CHF. A way of attenuating these undesired temperature excursions has been identified through the application of high enough initial heating power.

Study of reverse Brayton cryocooler with Helium-Neon mixture for HTS cable

Science.gov (United States)

Dhillon, A. K.; Ghosh, P.

2017-12-01

As observed in the earlier studies, helium is more efficient than neon as a refrigerant in a reverse Brayton cryocooler (RBC) from the thermodynamic point of view. However, the lower molecular weight of helium leads to higher refrigerant inventory as compared to neon. Thus, helium is suitable to realize the high thermodynamic efficiency of RBC whereas neon is appropriate for the compactness of the RBC. A binary mixture of helium and neon can be used to achieve high thermodynamic efficiency in the compact reverse Brayton cycle (RBC) based cryocooler. In this paper, an attempt has been made to analyze the thermodynamic performance of the RBC with a binary mixture of helium and neon as the working fluid to provide 1 kW cooling load for high temperature superconductor (HTS) power cables working with a temperature range of 50 K to 70 K. The basic RBC is simulated using Aspen HYSYS V8.6®, a commercial process simulator. Sizing of each component based on the optimized process parameters for each refrigerant is performed based on a computer code developed using Engineering Equation Solver (EES-V9.1). The recommendation is provided for the optimum mixture composition of the refrigerant based on the trade-off factors like thermodynamic efficiency such as the exergy efficiency and equipment considerations. The outcome of this study may be useful for recommending a suitable refrigerant for the RBC operating at a temperature level of 50 K to 70 K.

Water chemistry in nuclear power stations with high-temperature reactors with particular reference to the AVR

International Nuclear Information System (INIS)

Nieder, R.; Resch, G.

1976-01-01

The water-steam cycle of a nuclear power plant with a helium-cooled high-temperature reactor differs in design data significantly and extensively from the corresponding cycles of light-water-cooled nuclear reactors and resembles to a great extent the water-steamcycle of a modern conventional power plant. The radioactive constituents of the water-steam cycle can be satisfactorily removed apart from Tritium by means of a pre-coat filter with powder-resisn, as comprehensive experiments have demonstrated. (orig.) [de

Simulation of the fuzzy-smith control system for the high temperature gas cooled reactor

International Nuclear Information System (INIS)

Li Deheng; Xu Xiaolin; Zheng Jie; Guo Renjun; Zhang Guifen

1997-01-01

The Fuzzy-Smith pre-estimate controller to solve the control of the big delay system is developed, accompanied with the development of the mathematical model of the 10 MW high temperature gas cooled test reactor (HTR-10) and the design of its control system. The simulation results show the Fuzzy-Smith pre-estimate controller has the advantages of both fuzzy control and Smith pre-estimate controller; it has better compensation to the delay and better adaptability to the parameter change of the control object. So it is applicable to the design of the control system for the high temperature gas cooled reactor

Prospects of power conversion technology of direct-cycle helium gas turbine for MHTGR

International Nuclear Information System (INIS)

Li Yong; Zhang Zuoyi

1999-01-01

The modular high temperature gas cooled reactor (MHTGR) is a modern passively safe reactor. The reactor and helium gas turbine may be combined for high efficiency's power conversion, because MHTGR has high outlet temperature up to 950 degree C. Two different schemes are planed separately by USA and South Africa. the helium gas turbine methodologies adopted by them are mainly based on the developed heavy duty industrial and aviation gas turbine technology. The author introduces the differences of two technologies and some design issues in the design and manufacture. Moreover, the author conclude that directly coupling a closed Brayton cycle gas turbine concept to the passively safe MHTGR is the developing direction of MHTGR due to its efficiency which is much higher than that of using steam turbine