Nuclear design for high temperature gas cooled reactor (GTHTR300C) using MOX fuel

International Nuclear Information System (INIS)

Mouri, Tomoaki; Kunitomi, Kazuhiko

2008-01-01

A design study of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C) that can produce both electricity and hydrogen has been carried out in Japan Atomic Energy Agency. The GTHTR300C is the system with thermal power of 600MW and reactor outlet temperature of 950degC, which is expected to supply the hydrogen to fuel cell vehicles after 2020s. In future, the full deployment of fast reactor cycle without natural uranium will demand the use of Mixed-Oxide (MOX) fuels in the GTHTR300C. Therefore, a nuclear design was performed to confirm the feasibility of the reactor core using MOX fuels. The designed reactor core has high performance and meets safety requirements. In this paper, the outline of the GTHTR300C and the nuclear design of the reactor core using MOX fuels are described. (author)

Composite Materials under Extreme Radiation and Temperature Environments of the Next Generation Nuclear Reactors

International Nuclear Information System (INIS)

Simos, N.

2011-01-01

In the nuclear energy renaissance, driven by fission reactor concepts utilizing very high temperatures and fast neutron spectra, materials with enhanced performance that exceeds are expected to play a central role. With the operating temperatures of the Generation III reactors bringing the classical reactor materials close to their performance limits there is an urgent need to develop and qualify new alloys and composites. Efforts have been focused on the intricate relations and the high demands placed on materials at the anticipated extreme states within the next generation fusion and fission reactors which combine high radiation fluxes, elevated temperatures and aggressive environments. While nuclear reactors have been in operation for several decades, the structural materials associated with the next generation options need to endure much higher temperatures (1200 C), higher neutron doses (tens of displacements per atom, dpa), and extremely corrosive environments, which are beyond the experience on materials accumulated to-date. The most important consideration is the performance and reliability of structural materials for both in-core and out-of-core functions. While there exists a great body of nuclear materials research and operating experience/performance from fission reactors where epithermal and thermal neutrons interact with materials and alter their physio-mechanical properties, a process that is well understood by now, there are no operating or even experimental facilities that will facilitate the extreme conditions of flux and temperature anticipated and thus provide insights into the behaviour of these well understood materials. Materials, however, still need to be developed and their interaction and damage potential or lifetime to be quantified for the next generation nuclear energy. Based on material development advances, composites, and in particular ceramic composites, seem to inherently possess properties suitable for key functions within the

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

High temperature reactor and application to nuclear process heat

Energy Technology Data Exchange (ETDEWEB)

Schulten, R; Kugeler, K [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.)

1976-01-01

The principle of high temperature nuclear process heat is explained and the main applications (hydrogasification of coal, nuclear chemical heat pipe, direct reduction of iron ore, coal gasification by steam and water splitting) are described in more detail. The motivation for the introduction of nuclear process heat to the market, questions of cost, of raw material resources and environmental aspects are the next point of discussion. The new technological questions of the nuclear reactor and the status of development are described, especially information about the fuel elements, the hot gas ducts, the contamination and some design considerations are added. Furthermore the status of development of helium heated steam reformers, the main results of the work until now and the further activities in this field are explained.

Temperature conditions of foundation plates under nuclear power plant reactor compartments

International Nuclear Information System (INIS)

Ehsaulov, S.L.

1990-01-01

Method for calculation of temperature conditions for foundation plates under reactor compartments located in the main building, used in construction of the second stage of the Kostroma nuclear power plant, is considered. The obtained calculation data can be used for determining the most suitable period of concrete placement, composition, initial temperature, manufacturing technology and ways of delivery of concrete mixture

High-temperature nuclear reactor power plant cycle for hydrogen and electricity production – numerical analysis

Directory of Open Access Journals (Sweden)

Dudek Michał

2016-01-01

Full Text Available High temperature gas-cooled nuclear reactor (called HTR or HTGR for both electricity generation and hydrogen production is analysed. The HTR reactor because of the relatively high temperature of coolant could be combined with a steam or gas turbine, as well as with the system for heat delivery for high-temperature hydrogen production. However, the current development of HTR’s allows us to consider achievable working temperature up to 750°C. Due to this fact, industrial-scale hydrogen production using copper-chlorine (Cu-Cl thermochemical cycle is considered and compared with high-temperature electrolysis. Presented calculations show and confirm the potential of HTR’s as a future solution for hydrogen production without CO2 emission. Furthermore, integration of a hightemperature nuclear reactor with a combined cycle for electricity and hydrogen production may reach very high efficiency and could possibly lead to a significant decrease of hydrogen production costs.

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.)

Temperature distribution due to the heat generation in nuclear reactor shielding

International Nuclear Information System (INIS)

Torres, L.M.R.

1985-01-01

A study is performed for calculating nuclear heating due to the interaction of neutrons and gamma-rays with matter. Modifications were implemented in the ANISN and DOT 3.5 codes, that solve the transport equation using the discrete ordinate method, in one two-dimensions respectively, to include nuclear heating calculations in these codes. In order to determine the temperature distribution, using the finite difference method, a numerical model was developed for solving the heat conduction equation in one-dimension, in plane, cylindrical and spherical geometries, and in two-dimensions, X-Y and R-Z geometries. Based on these models, computer programs were developed for calculating the temperature distribution. Tests and applications of the implemented modifications were performed in problems of nuclear heating and temperature distribution due to radiation energy deposition in fission and fusion reactor shields. (Author) [pt

Nuclear rocket engine reactor

Energy Technology Data Exchange (ETDEWEB)

Lanin, Anatoly

2013-07-01

Covers a new technology of nuclear reactors and the related materials aspects. Integrates physics, materials science and engineering Serves as a basic book for nuclear engineers and nuclear physicists. The development of a nuclear rocket engine reactor (NRER) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

Neutron analysis of the fuel of high temperature nuclear reactors

International Nuclear Information System (INIS)

Bastida O, G. E.; Francois L, J. L.

2014-10-01

In this work a neutron analysis of the fuel of some high temperature nuclear reactors is presented, studying its main features, besides some alternatives of compound fuel by uranium and plutonium, and of coolant: sodium and helium. For this study was necessary the use of a code able to carry out a reliable calculation of the main parameters of the fuel. The use of the Monte Carlo method was convenient to simulate the neutrons transport in the reactor core, which is the base of the Serpent code, with which the calculations will be made for the analysis. (Author)

Nuclear reactor

International Nuclear Information System (INIS)

Mysels, K.J.; Shenoy, A.S.

1976-01-01

A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

Graphite materials for nuclear reactors

International Nuclear Information System (INIS)

Oku, Tatsuo

1991-01-01

Graphite materials have been used in the nuclear fission reactors from the beginning of the reactor development for the speed reduction and reflection of neutron. Graphite materials are used both as a moderator and as a reflector in the core of high temperature gas-cooled reactors, and both as a radiation shielding material and as a reflector in the surrounding of the core for the fast breeder reactor. On the other hand, graphite materials are being positively used as a first wall of plasma as it is known that low Z materials are useful for holding high temperature plasma in the nuclear fusion devices. In this paper the present status of the application of graphite materials to the nuclear fission reactors and fusion devices (reactors) is presented. In addition, a part of results on the related properties to the structural design and safety evaluation and results examined on the subjects that should be done in the future are also described. (author)

Small reactors for low-temperature nuclear heat applications

International Nuclear Information System (INIS)

1988-06-01

In accordance with the Member States' calls for information exchange in the field of nuclear heat application (NHA) two IAEA meetings were organized already in 1976 and 1977. After this ''promising period'', the development of relevant programmes in IAEA Member States was slowed down and therefore only after several years interruption a new Technical Committee Meeting with a Workshop was organized in late 1983, to review the status of NHA, after a few new specific plans appeared in some IAEA Member States in the early 1980's for the use of heat from existing or constructed NPPs and for developing nuclear heating plants (NHP). In June 1987 an Advisory Group Meeting was convened in Winnipeg, Canada, to discuss and formulate a state-of-the-art review on ''Small Reactors for Low Temperature Nuclear Heat Application''. Information on this subject gained up to 1987 in the Member States whose experts attended this meeting is embodied in the present Technical Report. Figs and tabs

The pebble-bed high-temperature reactor as a source of nuclear process heat. Vol. 3

International Nuclear Information System (INIS)

Kugeler, K.; Schulten, R.; Kugeler, M.; Niessen, H.F.; Roeth-Kamat, M.; Hohn, H.; Woike, O.; Germer, J.H.

1974-08-01

The characteristic questions concerning a process heat reactor with high helium outlet temperatures are dealt with in this volume like e.g. fuel element design, corrosion, and fission product release. Furthermore, some possibilities of the technical realization of the hot-gas ducting and intermediate heat exchangers are described. Important parameters for the design of the reactor such as core power density, helium inlet and outlet temperatures, helium pressure and fuel cycle burn-up and conversion and the effect of these on the primary circuit are investigated. The important question regarding which reactor vessel is to be chosen for nuclear process heat plants is discussed with the aid of the integrated and non-integrated concepts using prestressed concrete, cast iron and cast steel. Thereafter, considerations on the safety of the nuclear plant are given. Finally, mention is made of the availability of the nuclear plant and of the status of development of the HTR technology. (orig.) [de

Real-time reactor coolant system pressure/temperature limit system

International Nuclear Information System (INIS)

Newton, D.G.; Schemmel, R.R.; Van Scooter, W.E. Jr.

1991-01-01

This patent describes an system, used in controlling the operating of a nuclear reactor coolant system, which automatically calculates and displays allowable reactor coolant system pressure/temperature limits within the nuclear reactor coolant system based upon real-time inputs. It comprises: means for producing signals representative of real-time operating parameters of the nuclear reactor cooling system; means for developing pressure and temperature limits relating the real-time operating parameters of the nuclear reactor coolant system, for normal and emergency operation thereof; means for processing the signals representative of real-time operating parameters of the nuclear reactor coolant system to perform calculations of a best estimate of signals, check manual inputs against permissible valves and test data acquisition hardware for validity and over/under range; and means for comparing the representative signals with limits for the real-time operating parameters to produce a signal for a real-time display of the pressure and temperature limits and of the real-time operating parameters use an operator in controlling the operation of the nuclear reactor coolant system

Monte Carlo calculation of the nuclear temperature coefficient in fast reactors

Energy Technology Data Exchange (ETDEWEB)

Matthes, W.

1974-04-15

A Monte Carlo program for the calculation of the nuclear temperature coefficient for fast reactors is described. The special difficulties for this problem are the energy and space dependence of the cross sections and the calculation of differential eifects. These difficulties are discussed in detail and the way for their solution chosen in this program is described. (auth)

Economic analysis of multiple-module high temperature gas-cooled reactor (MHTR) nuclear power plants

International Nuclear Information System (INIS)

Liu Yu; Dong Yujie

2011-01-01

In recent years, as the increasing demand of energy all over the world, and the pressure on greenhouse emissions, there's a new opportunity for the development of nuclear energy. Modular High Temperature Gas-cooled Reactor (MHTR) received recognition for its inherent safety feature and high outlet temperature. Whether the Modular High Temperature Gas-cooled Reactor would be accepted extensively, its economy is a key point. In this paper, the methods of qualitative analysis and the method of quantitative analysis, the economic models designed by Economic Modeling Working Group (EMWG) of the Generation IV International Forum (GIF), as well as the HTR-PM's main technical features, are used to analyze the economy of the MHTR. A prediction is made on the basis of summarizing High Temperature Gas-cooled Reactor module characteristics, construction cost, total capital cost, fuel cost and operation and maintenance (O and M) cost and so on. In the following part, comparative analysis is taken measures to the economy and cost ratio of different designs, to explore the impacts of modularization and standardization on the construction of multiple-module reactor nuclear power plant. Meanwhile, the analysis is also adopted in the research of key factors such as the learning effect and yield to find out their impacts on the large scale development of MHTR. Furthermore, some reference would be provided to its wide application based on these analysis. (author)

The modular high-temperature gas-cooled reactor: A cost/risk competitive nuclear option

International Nuclear Information System (INIS)

Gotschall, H.L.

1994-01-01

The business risks of nuclear plant ownership are identified as a constraint on the expanded use of nuclear power. Such risks stem from the exacting demands placed on owner/operator organizations of current plants to demonstrate ongoing compliance with safety regulations and the resulting high costs for operation and maintenance. This paper describes the Modular High-Temperature Gas-Cooled Reactor (MHTGR) design, competitive economics, and approach to reducing the business risks of nuclear plant ownership

Neutron analysis of the fuel of high temperature nuclear reactors; Analisis neutronico del combustible de reactores nucleares de alta temperatura

Energy Technology Data Exchange (ETDEWEB)

Bastida O, G. E.; Francois L, J. L., E-mail: gbo729@yahoo.com.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

2014-10-15

In this work a neutron analysis of the fuel of some high temperature nuclear reactors is presented, studying its main features, besides some alternatives of compound fuel by uranium and plutonium, and of coolant: sodium and helium. For this study was necessary the use of a code able to carry out a reliable calculation of the main parameters of the fuel. The use of the Monte Carlo method was convenient to simulate the neutrons transport in the reactor core, which is the base of the Serpent code, with which the calculations will be made for the analysis. (Author)

Oklo reactors and implications for nuclear science

OpenAIRE

Davis, E. D.; Gould, C. R.; Sharapov, E. I.

2014-01-01

We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlie...

Nuclear reactor development in China for non-electrical applications

International Nuclear Information System (INIS)

Sun Yuliang; Zhong Daxin; Dong Duo; Xu Yuanhui

1998-01-01

In parallel to its vigorous program of nuclear power generation, China has attached great importance to the development of nuclear reactors for non-electrical applications. The Institute of Nuclear Energy Technology (INET) in Beijing has been developing technologies of the water-cooled heating reactor and the modular high temperature gas-cooled reactor. In 1989, a 5 MW water cooled test reactor was erected. Currently, an industrial demonstration nuclear heating plant is being projected. Feasibility studies are being made of sea-water desalination using the INET developed nuclear heating reactor as heat source. Also, a 10 MW high temperature gas-cooled test reactor is being constructed at INET in the framework of China's national high-tech program. The paper gives an overview of China's energy market situation. With respect to China's technology development of high temperature gas-cooled reactors and water cooled heating reactors, the paper describes some general requirements on the technical development, reviews the national programs and activities, describes briefly the design and safety features of the reactor concepts, discusses aspects of application potentials. (author)

Nuclear research reactors in Brazil

Energy Technology Data Exchange (ETDEWEB)

Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

Nuclear research reactors in Brazil

International Nuclear Information System (INIS)

Cota, Anna Paula Leite; Mesquita, Amir Zacarias

2011-01-01

The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

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.)

Parametric Evaluation of Large-Scale High-Temperature Electrolysis Hydrogen Production Using Different Advanced Nuclear Reactor Heat Sources

International Nuclear Information System (INIS)

Harvego, Edwin A.; McKellar, Michael G.; O'Brien, James E.; Herring, J. Stephen

2009-01-01

High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 C to 950 C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycles producing the highest efficiencies varied depending on the temperature range considered

Nuclear reactor

International Nuclear Information System (INIS)

Tilliette, Z.

1975-01-01

A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

Method for operating nuclear reactor

International Nuclear Information System (INIS)

Utamura, Motoaki; Urata, Megumu; Uchida, Shunsuke

1978-01-01

Purpose: In order to judge the fuel failures, if any, without opening a reactor container for BWR type reactors, a method has been described for measuring the difference between the temperature dependent iodine spike value and the pressure dependent iodine spike value in the pressure vessel. Method: After the scram of a nuclear reactor, steam generated by decay heat is condensed in a remaining heat exchanger and cooling water is returned through a recycling pipe line to a reactor core. At the same time, a control rod drive system pump is operated, the reactor core is filled with the cooling water. Then, the coolant is taken from the recycling pipe line to cool the reactor core. After applying the temperature fluctuation, the cooling water is sampled at a predetermined time interval at a sampling point to determine the changes with time in the radioactive concentration of iodine. When the radioactivity of iodine in the cooling water is lowered sufficiently by a reactor purifying system, the nuclear reactor vessel is depressurized. After applying pressure fluctuation, iodine spike value is determined. (Kawakami, Y.)

Advanced Intermediate Heat Transport Loop Design Configurations for Hydrogen Production Using High Temperature Nuclear Reactors

International Nuclear Information System (INIS)

Chang Oh; Cliff Davis; Rober Barner; Paul Pickard

2005-01-01

The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. A number of possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermal-hydraulic evaluations and cycle-efficiency evaluations of the different configurations and coolants. The thermal-hydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various

Nuclear reactors

International Nuclear Information System (INIS)

Prescott, R.F.

1976-01-01

A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates

Hydrogen production by high temperature electrolysis of water vapour and nuclear reactors

International Nuclear Information System (INIS)

Jean-Pierre Py; Alain Capitaine

2006-01-01

This paper presents hydrogen production by a nuclear reactor (High Temperature Reactor, HTR or Pressurized Water Reactor, PWR) coupled to a High Temperature Electrolyser (HTE) plant. With respect to the coupling of a HTR with a HTE plant, EDF and AREVA NP had previously selected a combined cycle HTR scheme to convert the reactor heat into electricity. In that case, the steam required for the electrolyser plant is provided either directly from the steam turbine cycle or from a heat exchanger connected with such cycle. Hydrogen efficiency production is valued using high temperature electrolysis. Electrolysis production of hydrogen can be performed with significantly higher thermal efficiencies by operating in the steam phase than in the water phase. The electrolysis performance is assessed with solid oxide and solid proton electrolysis cells. The efficiency from the three operating conditions (endo-thermal, auto-thermal and thermo-neutral) of a high temperature electrolysis process is evaluated. The technical difficulties to use the gases enthalpy to heat the water are analyzed, taking into account efficiency and technological challenges. EDF and AREVA NP have performed an analysis to select an optimized process giving consideration to plant efficiency, plant operation, investment and production costs. The paper provides pathways and identifies R and D actions to reach hydrogen production costs competitive with those of other hydrogen production processes. (authors)

Safety of nuclear reactors - Part A - unsteady state temperature history mathematical model

International Nuclear Information System (INIS)

El-Shayeb, M.; Yusoff, M.Z.; Boosroh, M.H.; Ideris, F.; Hasmady Abu Hassan, S.; Bondok, A.

2004-01-01

A nuclear reactor structure under abnormal operations of near meltdown will be exposed to a tremendous amount of heat flux in addition to the stress field applied under normal operation. Temperature encountered in such case is assumed to be beyond 1000 Celsius degrees. A 2-dimensional mathematical model based on finite difference methods, has been developed for the fire resistance calculation of a concrete-filled square steel column with respect to its temperature history. Effects due to nuclear radiation and mechanical vibrations will be explored in a later future model. The temperature rise in each element can be derived from its heat balance by applying the parabolic unsteady state, partial differential equation and numerical solution into the steel region. Calculation of the temperature of the elementary regions needs to satisfy the symmetry conditions and the relevant material properties. The developed mathematical model is capable to predict the temperature history in the column and on the surface with respect to time. (authors)

Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

Science.gov (United States)

Bowles, K. J.

1974-01-01

An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

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

Transmutation of nuclear waste in nuclear reactors

International Nuclear Information System (INIS)

Abrahams, K.; Kloosterman, J.L.; Pilate, S.; Wehmann, U.K.

1996-03-01

The objective of this joint study of ECN, Belgonucleaire, and Siemens is to investigate possibilities for transmutation of nuclear waste in regular nuclear reactors or in special transmutation devices. Studies of possibilities included the limits and technological development steps which would be needed. Burning plutonium in fast reactors, gas-cooled high-temperature reactors and light water reactors (LWR) have been considered. For minor actinides the transmutation rate mainly depends on the content of the minor actinides in the reactor and to a much less degree on the fact whether one uses a homogeneous system (with the actinides mixed into the fuel) or a heterogeneous system. If one wishes to stabilise the amount of actinides from the present LWRs, about 20% of all nuclear power would have to be generated in special burner reactors. It turned out that reactor transmutation of fission products would require considerable recycling efforts and that the time needed for a substantial transmutation would be rather long for the presently available levels of the neutron flux. If one would like to design burner systems which can serve more light water reactors, a large effort would be needed and other burners (possibly driven by accelerators) should be considered. (orig.)

Cost comparison of very high temperature nuclear reactors for process heat applications

International Nuclear Information System (INIS)

Crowley, J.H.; Newman, J.B.

1975-03-01

In April 1974, the United States Atomic Energy Commission (USAEC) authorized General Atomic Company, General Electric Company and Westinghouse Astronuclear Laboratory to assess the available technology for producing process heat utilizing very high temperature nuclear reactors. General Electric and Westinghouse produced concepts for the entire nuclear system, including the balance of plant. The General Atomic assessment included only the nuclear reactor portion of the nuclear plant. United Engineers and Constructors Inc. (UE and C) was requested by the USAEC in November 1974 to prepare an economic comparison of the three conceptual plants. The comparison is divided into three tasks: (1) Develop a balance of plant conceptual design to be combined with the General Atomic concept as a basis for comparison, and estimate the cost of the General Atomic/UE and C concept in July 1974 dollars; (2) Normalize the overall plant costs for the General Atomic/UE and C, General Electric and Westinghouse concepts, compare the costs, and identify significant differences between the concepts; and (3) Estimate the operation and maintenance costs for the General Atomic/UE and C plant and compare with the other concepts. The results of these task studies are discussed

HOMOGENEOUS NUCLEAR POWER REACTOR

Science.gov (United States)

King, L.D.P.

1959-09-01

A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

Assessment of the Capability of Molten Salt Reactors as a Next Generation High Temperature Reactors

International Nuclear Information System (INIS)

Elsheikh, B.M.

2017-01-01

Molten Salt Reactor according to Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment (MSRE) programs, was designed to be the first full-scale, commercial nuclear power plant utilizing molten salt liquid fuels that can be used for producing electricity, and producing fissile fuels (breeding)burning actinides. The high temperature in the primary cycle enables the realization of efficient thermal conversion cycles with net thermal efficiencies reach in some of the designs of nuclear reactors greater than 45%. Molten salts and liquid salt because of their low vapor pressure are excellent candidates for meeting most of the requirements of these high temperature reactors. There is renewed interest in MSRs because of changing goals and new technologies in the use of high-temperature reactors. Molten Salt Reactors for high temperature create substantial technical challenges to have high effectiveness intermediate heat transfer loop components. This paper will discuss and investigate the capability and compatibility of molten salt reactors, toward next generation high temperature energy system and its technical challenges

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.

Inherent safe design of advanced high temperature reactors - concepts for future nuclear power plants

International Nuclear Information System (INIS)

Hodzic, A.; Kugeler, K.

1997-01-01

This paper discusses the applicable solutions for a commercial size High Temperature Reactor (HTR) with inherent safety features. It describes the possible realization using an advanced concept which combines newly proposed design characteristics with some well known and proven HTR inherent safety features. The use of the HTR technology offers the conceivably best solution to meet the legal criteria, recently stated in Germany, for the future reactor generation. Both systems, block and pebble bed ,reactor, could be under certain design conditions self regulating in terms of core nuclear heat, mechanical stability and the environmental transfer. 23 refs., 7 figs

Nuclear reactors to come

International Nuclear Information System (INIS)

Lung, M.

2002-01-01

The demand for nuclear energy will continue to grow at least till 2050 because of mainly 6 reasons: 1) the steady increase of the world population, 2) China, India and Indonesia will reach higher social standard and their energy consumption will consequently grow, 3) fossil energy resources are dwindling, 4) coal will be little by little banned because of its major contribution to the emission of green house effect gas, 5) renewable energies need important technological jumps to be really efficient and to take the lead, and 6) fusion energy is not yet ready to take over. All these reasons draw a promising future for nuclear energy. Today 450 nuclear reactors are operating throughout the world producing 17% of the total electrical power demand. In order to benefit fully of this future, nuclear industry has to improve some characteristics of reactors: 1) a more efficient use of uranium (it means higher burnups), 2) a simplification and automation of reprocessing-recycling chain of processes, 3) efficient measures against proliferation and against any misuse for terrorist purposes, and 4) an enhancement of safety for the next generation of reactors. The characteristics of fast reactors and of high-temperature reactors will likely make these kinds of reactors the best tools for energy production in the second half of this century. (A.C.)

Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

Energy Technology Data Exchange (ETDEWEB)

Ramirez S, J. R.; Alonso V, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2013-10-15

In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

Very high temperature measurements: Applications to nuclear reactor safety tests

International Nuclear Information System (INIS)

Parga, Clemente-Jose

2013-01-01

This PhD dissertation focuses on the improvement of very high temperature thermometry (1100 deg. C to 2480 deg. C), with special emphasis on the application to the field of nuclear reactor safety and severe accident research. Two main projects were undertaken to achieve this objective: - The development, testing and transposition of high-temperature fixed point (HTFP) metal-carbon eutectic cells, from metrology laboratory precision (±0.001 deg. C) to applied research with a reasonable degradation of uncertainties (±3-5 deg. C). - The corrosion study and metallurgical characterization of Type-C thermocouple (service temp. 2300 deg. C) prospective sheath material was undertaken to extend the survivability of TCs used for molten metallic/oxide corium thermometry (below 2000 deg. C)

Nuclear reactor design

CERN Document Server

2014-01-01

This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

Summary - Advanced high-temperature reactor for hydrogen and electricity production

International Nuclear Information System (INIS)

Forsberg, Charles W.

2001-01-01

Historically, the production of electricity has been assumed to be the primary application of nuclear energy. That may change. The production of hydrogen (H 2 ) may become a significant application. The technology to produce H 2 using nuclear energy imposes different requirements on the reactor, which, in turn, may require development of new types of reactors. Advanced High Temperature reactors can meet the high temperature requirements to achieve this goal. This alternative application of nuclear energy may necessitate changes in the regulatory structure

On the selfacting safe limitation of fission power and fuel temperature in innovative nuclear reactors

International Nuclear Information System (INIS)

Scherer, W.; Brockmann, H.; Drecker, S.; Gerwin, H.; Haas, K.A.; Kugeler, K.; Ohlig, U.; Ruetten, H.J.; Teuchert, E.; Werner, H.; Wolf, L.

1994-08-01

Nuclear energy probably will not contribute significantly to the future worldwide energy supply until it can be made catastrophe-free. Therefore it has to be shown, that the consequences of even largest accidents will have no major impact to the environment of a power plant. In this paper one of the basic conditions for such a nuclear technology is discussed. Using mainly the modular pebble-bed high-temperature reactor as an example, the design principles, analytical methods and the level of knowledge as given today in controlling reactivity accidents by inherent safety features of innovative nuclear reactors are described. Complementary possibilities are shown to reach this goal with systems of different types of construction. Questions open today and resulting requirements for future activities are discussed. Today's knowledge credibly supports the possibility of a catastrophe-free nuclear technology with respect to reactivity events. (orig.)

Nuclear Reactor Physics

Science.gov (United States)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D

International Nuclear Information System (INIS)

Scari, Maria Elizabeth

2017-01-01

Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF 2 , the LiF-BeF 2 , also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO (Tristructural

Selection of nuclear reactor coolant materials

International Nuclear Information System (INIS)

Shi Lisheng; Wang Bairong

2012-01-01

Nuclear material is nuclear material or materials used in nuclear industry, the general term, it is the material basis for the construction of nuclear power, but also a leader in nuclear energy development, the two interdependent and mutually reinforcing. At the same time, nuclear materials research, development and application of the depth and breadth of science and technology reflects a nation and the level of the nuclear power industry. Coolant also known as heat-carrier agent, is an important part of the heart nuclear reactor, its role is to secure as much as possible to the economic output in the form fission energy to heat the reactor to be used: the same time cooling the core, is controlled by the various structural components allowable temperature. This paper described the definition of nuclear reactor coolant and characteristics, and then addressed the requirements of the coolant material, and finally were introduced several useful properties of the coolant and chemical control. (authors)

Synergistic smart fuel for in-pile nuclear reactor measurements

Energy Technology Data Exchange (ETDEWEB)

Smith, J.A.; Kotter, D.K. [Idaho National Laboratories, Idaho Falls (United States); Ali, R.A.; Garrett, S.L. [Penn State University, University Park, State College, PA 16801 (United States)

2013-07-01

The thermo-acoustic fuel rod sensor developed in this research has demonstrated a novel technique for monitoring the temperature within the core of a nuclear reactor or the temperature of the surrounding heat-transfer fluid. It uses the heat from the nuclear fuel to generate sustained acoustic oscillations whose frequency will be indicative of the temperature. Converting a nuclear fuel rod into this type of thermo-acoustic sensor simply requires the insertion of a porous material (stack). This sensor has demonstrated a synergy with the elevated temperatures that exist within the nuclear reactor using materials that have only minimal susceptibility to high-energy particle fluxes. When the sensor is in operation, the sound waves radiated from the fuel rod resonator will propagate through the surrounding cooling fluid. The frequency of these oscillations is directly correlated with an effective temperature within the fuel rod resonator. This device is self-powered and is operational even in case of total loss of power of the reactor.

Nuclear reactor shutdown control rod assembly

International Nuclear Information System (INIS)

Bilibin, K.

1988-01-01

This patent describes a nuclear reactor having a reactor core and a reactor coolant flowing therethrough, a temperature responsive, self-actuated nuclear reactor shutdown control rod assembly, comprising: an upper drive line terminating at its lower end with a substantially cylindrical wall member having inner and outer surfaces; a lower drive line having a lower end adapted to be attached to a neutron absorber; a ring movable disposed about the outer surface of the wall member of the upper drive line; thermal actuation means adapted to be in heat exchange relationship with coolant in an associated reactor core and in contact with the ring, and balls located within the openings in the upper drive line. When reactor coolant approaches a predetermined design temperature the actuation means moves the ring sufficiently so that the balls move radially out from the recess and into the space formed by the second portion of the ring thereby removing the vertical support for the lower drive line such that the lower drive line moves downwardly and inserts an associated neutron absorber into an associated reactor core resulting in automatic reduction of reactor power

Passive cooling of a fixed bed nuclear reactor

International Nuclear Information System (INIS)

Petry, V.J.; Bortoli, A.L. de; Sefidwash, F.

2005-01-01

Small nuclear reactors without the need for on-site refuelling have greater simplicity, better compliance with passive safety systems, and are more adequate for countries with small electric grids and limited investment capabilities. Here the passive cooling characteristic of the fixed bed nuclear reactor (FBNR), that is being developed under the International Atomic Energy Agency (IAEA) Coordinated Research Project, is studied. A mathematical model is developed to calculate the temperature distribution in the fuel chamber of the reactor. The results demonstrate the passive cooling of this nuclear reactor concept. (authors)

High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

International Nuclear Information System (INIS)

Beck, J.M.; Collins, J.W.; Garcia, C.B.; Pincock, L.F.

2010-01-01

High Temperature Gas Reactors (HTGR) have been designed and operated throughout the world over the past five decades. These seven HTGRs are varied in size, outlet temperature, primary fluid, and purpose. However, there is much the Next Generation Nuclear Plant (NGNP) has learned and can learn from these experiences. This report captures these various experiences and documents the lessons learned according to the physical NGNP hardware (i.e., systems, subsystems, and components) affected thereby.

Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen Production Plant

International Nuclear Information System (INIS)

E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

2008-01-01

A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540 C and 900 C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating-current, AC, to direct-current, DC, conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%

Temperature measurements inside nuclear reactor cores

International Nuclear Information System (INIS)

Tarassenko, Serge

1969-11-01

Non negligible errors may happen in nuclear reactor temperature measurements using magnesium oxide insulated and stainless steel sheathed micro-wire thermocouples, when these thermometric lines are placed under operational conditions typical of electrical power stations. The present work shows that this error is principally due to electrical hysteresis and polarization phenomena in the insulator subjected to the strong fields generated by common-mode voltages. These phenomena favour the unsymmetrical common-mode current flow and thus lead to the differential-mode voltage generation which is superposing on the thermoelectric hot junction potential. A calculation and an experimental approach make possible the importance of the magnesium oxide insulating characteristics, the hot junction insulation, the choice of the main circuits in the data processing equipment as well as the galvanic isolation performances and the common-mode rejection features of all the measurement circuits. A justification is thereby given for the severe conditions imposed for the acceptance of thermoelectric materials; some particular precautions to be taken are described, as well as the high performance characteristics which have to be taken into account in choosing measurement systems linked to thermometric circuits with sheathed micro-wire thermocouples. (author) [fr

Combined conditioning in the high-temperature experimental nuclear reactor (AVR) at Juelich

International Nuclear Information System (INIS)

Nieder, R.; Vey, K.; Ivens, G.

1984-01-01

The high temperature experimental nuclear reactor (AVR) is the first nuclear power plant in which combined cycle operation has been introduced. The water-steam cycle has been operated for about 15 years according to the alkali method of working with ammonia and hydrazine. The VGB-guidelines have been adhered to througout. Since January 1983 cobined cycle operation has been employed, and in this process a pH-value of about 8.5 and an oxygen concentration of about 200 μg/kg in the feedwater have been used. A distinct reduction of tritium concentration in the water-steam cycle was the outstanding new result. (orig.) [de

Nuclear Reactor/Hydrogen Process Interface Including the HyPEP Model

International Nuclear Information System (INIS)

Steven R. Sherman

2007-01-01

The Nuclear Reactor/Hydrogen Plant interface is the intermediate heat transport loop that will connect a very high temperature gas-cooled nuclear reactor (VHTR) to a thermochemical, high-temperature electrolysis, or hybrid hydrogen production plant. A prototype plant called the Next Generation Nuclear Plant (NGNP) is planned for construction and operation at the Idaho National Laboratory in the 2018-2021 timeframe, and will involve a VHTR, a high-temperature interface, and a hydrogen production plant. The interface is responsible for transporting high-temperature thermal energy from the nuclear reactor to the hydrogen production plant while protecting the nuclear plant from operational disturbances at the hydrogen plant. Development of the interface is occurring under the DOE Nuclear Hydrogen Initiative (NHI) and involves the study, design, and development of high-temperature heat exchangers, heat transport systems, materials, safety, and integrated system models. Research and development work on the system interface began in 2004 and is expected to continue at least until the start of construction of an engineering-scale demonstration plant

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

Isothermal temperature reactivity coefficient measurement in TRIGA reactor

International Nuclear Information System (INIS)

Zagar, T.; Ravnik, M.; Trkov, A.

2002-01-01

Direct measurement of an isothermal temperature reactivity coefficient at room temperatures in TRIGA Mark II research reactor at Jozef Stefan Institute in Ljubljana is presented. Temperature reactivity coefficient was measured in the temperature range between 15 o C and 25 o C. All reactivity measurements were performed at almost zero reactor power to reduce or completely eliminate nuclear heating. Slow and steady temperature decrease was controlled using the reactor tank cooling system. In this way the temperatures of fuel, of moderator and of coolant were kept in equilibrium throughout the measurements. It was found out that TRIGA reactor core loaded with standard fuel elements with stainless steel cladding has small positive isothermal temperature reactivity coefficient in this temperature range.(author)

CFD Analysis of the Fuel Temperature in High Temperature Gas-Cooled Reactors

International Nuclear Information System (INIS)

In, W. K.; Chun, T. H.; Lee, W. J.; Chang, J. H.

2005-01-01

High temperature gas-cooled reactors (HTGR) have received a renewed interest as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor (PBR) and a prismatic modular reactor (PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both PBR and PMR. The objective of this study is to predict the fuel temperature distributions in PBR and PMR using a computational fluid dynamics(CFD) code, CFX-5. The reference reactor designs used in this analysis are PBMR400 and GT-MHR600

Nuclear reactor physics

CERN Document Server

Stacey, Weston M

2010-01-01

Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

Analysis of ultrasound propagation in high-temperature nuclear reactor feedwater to investigate a clamp-on ultrasonic pulse doppler flowmeter

International Nuclear Information System (INIS)

Tezuka, Kenichi; Mori, Michitsugu; Wada, Sanehiro; Aritomi, Masanori; Kikura, Hiroshige; Sakai, Yukihiro

2008-01-01

The flow rate of nuclear reactor feedwater is an important factor in the operation of a nuclear power reactor. Venturi nozzles are widely used to measure the flow rate. Other types of flowmeters have been proposed to improve measurement accuracy and permit the flow rate and reactor power to be increased. The ultrasonic pulse Doppler system is expected to be a candidate method because it can measure the flow profile across the pipe cross section, which changes with time. For accurate estimation of the flow velocity, the incidence angle of ultrasound entering the fluid should be estimated using Snell's law. However, evaluation of the ultrasound propagation is not straightforward, especially for a high-temperature pipe with a clamp-on ultrasonic Doppler flowmeter. The ultrasound beam path may differ from what is expected from Snell's law due to the temperature gradient in the wedge and variation in the acoustic impedance between interfaces. Recently, simulation code for ultrasound propagation has come into use in the nuclear field for nondestructive testing. This article analyzes and discusses ultrasound propagation, using 3D-FEM simulation code plus the Kirchhoff method, as it relates to flow profile measurement in nuclear reactor feedwater with the ultrasonic pulse Doppler system. (author)

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Butterfield, C.E.; Waite, E.

1982-01-01

A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

Development and testing of nuclear graphite for the German pebble-bed high temperature reactor

International Nuclear Information System (INIS)

Haag, G.; Delle, W.; Nickel, H.; Theymann, W.; Wilhelmi, G.

1987-01-01

Several types of high temperature reactors have been developed in the Federal Republic of Germany. They are all based on spherical fuel elements being surrounded by graphite as reflector material. As an example, HTR-500 developed by the Hochtemperatur Reaktorbau GmbH is shown. The core consists of the top reflector, the side reflector with inner and outer parts, the bottom reflector and the core support columns. The most serious problem with respect to fast neutron radiation damage had to be solved for the materials of those parts near the pebble bed. Regarding the temperature profile in the core, the top reflector is at 300 deg C, and as cooling gas flows from the top downward, the temperature of the inner side reflector rises to about 700 deg C at the bottom. Fortunately, the highest fast neutron load accumulated during the life time of a reactor corresponds to the lowest temperature. This makes graphite components easier to survive neutron exposure without being mechanically damaged, although the maximum fast neutron fluence is as high as 4 x 10 22 /cm 2 at about 400 deg C. HTR graphite components are divided into four classes according to loading. The raw materials for nuclear graphite, the development of pitch coke nuclear graphite, the irradiation behavior of ATR-2E and ASR-IRS and others are reported. (Kako, I.)

Virtual nuclear reactor for education of nuclear reactor physics

International Nuclear Information System (INIS)

Tsuji, Masashi; Narabayashi, Takashi; Shimazu, Youichiro

2008-01-01

As one of projects that were programmed in the cultivation program for human resources in nuclear engineering sponsored by the Ministry of Economy, Trade and Industry, the development of a virtual reactor for education of nuclear reactor physics started in 2007. The purpose of the virtual nuclear reactor is to make nuclear reactor physics easily understood with aid of visualization. In the first year of this project, the neutron slowing down process was visualized. The data needed for visualization are provided by Monte Carlo calculations; The flights of the respective neutrons generated by nuclear fissions are traced through a reactor core until they disappear by neutron absorption or slow down to a thermal energy. With this visualization and an attached supplement textbook, it is expected that the learners can learn more clearly the physical implication of neutron slowing process that is mathematically described by the Boltzmann neutron transport equation. (author)

HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

Science.gov (United States)

Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

1960-06-14

A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

Large-scale hydrogen production using nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Ryland, D.; Stolberg, L.; Kettner, A.; Gnanapragasam, N.; Suppiah, S. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

For many years, Atomic Energy of Canada Limited (AECL) has been studying the feasibility of using nuclear reactors, such as the Supercritical Water-cooled Reactor, as an energy source for large scale hydrogen production processes such as High Temperature Steam Electrolysis and the Copper-Chlorine thermochemical cycle. Recent progress includes the augmentation of AECL's experimental capabilities by the construction of experimental systems to test high temperature steam electrolysis button cells at ambient pressure and temperatures up to 850{sup o}C and CuCl/HCl electrolysis cells at pressures up to 7 bar and temperatures up to 100{sup o}C. In parallel, detailed models of solid oxide electrolysis cells and the CuCl/HCl electrolysis cell are being refined and validated using experimental data. Process models are also under development to assess options for economic integration of these hydrogen production processes with nuclear reactors. Options for large-scale energy storage, including hydrogen storage, are also under study. (author)

SCW Pressure-Channel Nuclear Reactor Some Design Features

Science.gov (United States)

Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

An investigation of temperature measurement methods in nuclear power plant reactor pressure vessel annealing

International Nuclear Information System (INIS)

Acton, R.U.; Gill, W.; Sais, D.J.; Schulze, D.H.; Nakos, J.T.

1996-05-01

The objective of this project was to provide an assessment of several methods by which the temperature of a commercial nuclear power plant reactor pressure vessel (RPV) could be measured during an annealing process. This project was a coordinated effort between DOE's Office of Nuclear Energy, Science and Technology; DOE's Light Water Reactor Technology Center at Sandia National Laboratories; and the Electric Power Research Institute's Non- Destructive Evaluation Center. Ball- thermocouple probes similar to those described in NUREG/CR-5760, spring-loaded, metal- sheathed thermocouple probes, and 1778 air- suspended thermocouples were investigated in experiments that heated a section of an RPV wall to simulate a thermal annealing treatment. A parametric study of ball material, emissivity, thermal conductivity, and thermocouple function locations was conducted. Also investigated was a sheathed thermocouple failure mode known as shunting (electrical breakdown of insulation separating the thermocouple wires). Large errors were found between the temperature as measured by the probes and the true RPV wall temperature during heat-up and cool-down. At the annealing soak temperature, in this case 454 degrees C [850'F], all sensors measured the same temperature within about ±5% (23.6 degrees C [42.5 degrees F]). Because of these errors, actual RPV wall heating and cooling rates differed from those prescribed (by up to 29%). Shunting does not appear to be a problem under these conditions. The large temperature measurement errors led to the development of a thermal model that predicts the RPV wall temperature from the temperature of a ball- probe. Comparisons between the model and the experimental data for ball-probes indicate that the model could be a useful tool in predicting the actual RPV temperature based on the indicated ball- probe temperature. The model does not predict the temperature as well for the spring-loaded and air suspended probes

Problems and prospects connected with development of high-temperature filtration technology at nuclear power plants equipped with VVER-1000 reactors

Science.gov (United States)

Shchelik, S. V.; Pavlov, A. S.

2013-07-01

Results of work on restoring the service properties of filtering material used in the high-temperature reactor coolant purification system of a VVER-1000 reactor are presented. A quantitative assessment is given to the effect from subjecting a high-temperature sorbent to backwashing operations carried out with the use of regular capacities available in the design process circuit in the first years of operation of Unit 3 at the Kalinin nuclear power plant. Approaches to optimizing this process are suggested. A conceptual idea about comprehensively solving the problem of achieving more efficient and safe operation of the high-temperature active water treatment system (AWT-1) on a nuclear power industry-wide scale is outlined.

Advanced nuclear reactor and nuclear fusion power generation

International Nuclear Information System (INIS)

2000-04-01

This book comprised of two issues. The first one is a advanced nuclear reactor which describes nuclear fuel cycle and advanced nuclear reactor like liquid-metal reactor, advanced converter, HTR and extra advanced nuclear reactors. The second one is nuclear fusion for generation energy, which explains practical conditions for nuclear fusion, principle of multiple magnetic field, current situation of research on nuclear fusion, conception for nuclear fusion reactor and economics on nuclear fusion reactor.

Materials for generation-IV nuclear reactors

International Nuclear Information System (INIS)

Alvarez, M. G.

2009-01-01

Materials science and materials development are key issues for the implementation of innovative reactor systems such as those defined in the framework of the Generation IV. Six systems have been selected for Generation IV consideration: gas-cooled fast reactor, lead-cooled fast reactor, molten salt-cooled reactor, sodium-cooled fast reactor, supercritical water-cooled reactor, and very high temperature reactor. The structural materials need to resist much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. For this reason, the first consideration in the development of Generation-IV concepts is selection and deployment of materials that operate successfully in the aggressive operating environments expected in the Gen-IV concepts. This paper summarizes the Gen-IV operating environments and describes the various candidate materials under consideration for use in different structural applications. (author)

Moderator for nuclear reactor

International Nuclear Information System (INIS)

Milgram, M.S.; Dunn, J.T.; Hart, R.S.

1995-01-01

This invention relates to a moderator for a nuclear reactor and more specifically, to a composite moderator. A moderator is designed to slow down, or thermalize, neutrons which are released during nuclear reactions in the reactor fuel. Pure or almost pure materials like light water, heavy water, beryllium or graphite are used singly as moderators at present. All these materials, are used widely. Graphite has a good mechanical strength at high temperatures encountered in the nuclear core and therefore is used as both the moderator and core structural material. It also exhibits a low neutron-capture cross section and high neutron scattering cross section. However, graphite is susceptible to attach by carbon dioxide and/or oxygen where applicable, and releases stress energy under certain circumstances, although under normal operating conditions these reactions can be controlled. (author). 1 tab

Desalination of seawater with nuclear reactors

International Nuclear Information System (INIS)

Nisan, S.; Volpi, L.

2003-01-01

About 40 % of the world population is concerned with water scarcity. This article reviews the different techniques of desalination: distillation (MED and MSF), reverse osmosis (RO), and electrodialysis (ED). The use of nuclear energy rests on several arguments: 1) it is economically efficient compared to fossil energy. 2) nuclear reactors provide heat covering a broad range of temperature, which allows the implementation of all the desalination techniques. 3) the heat normally lost at the heat sink could be used for desalination. And 4) nuclear is respectful of the environment. The feedback experience concerning nuclear desalination is estimated to about 100 reactor-years, it is sufficient to allow the understanding of all the physical and technological processes involved. In Japan, 8 PWR-type reactors are coupled to MED, MSF, and RO desalination techniques, the water produced is used locally mainly for feeding steam generators. (A.C.)

Desalination of seawater with nuclear reactors

International Nuclear Information System (INIS)

Nisan, S.; Volpi, L.

2001-01-01

About 40 % of the world population is concerned with water scarcity. This article reviews the different techniques of desalination: distillation (MED and MSF), reverse osmosis (RO), and electrodialysis (ED). The use of nuclear energy rests on several arguments: 1) it is economically efficient compared to fossil energy; 2) nuclear reactors provide heat covering a broad range of temperature, which allows the implementation of all the desalination techniques; 3) the heat normally lost at the heat sink could be used for desalination; and 4) nuclear is respectful of the environment. The feedback experience concerning nuclear desalination is estimated to about 100 reactor-years, it is sufficient to allow the understanding of all the physical and technological processes involved. In Japan, 8 PWR-type reactors are coupled to MED, MSF, and RO desalination techniques, the water produced is used locally mainly for feeding steam generators. (A.C.)

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

Nuclear research reactors

International Nuclear Information System (INIS)

1985-01-01

It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

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)

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)

Estimation of temperature distribution in a reactor shield

International Nuclear Information System (INIS)

Agarwal, R.A.; Goverdhan, P.; Gupta, S.K.

1989-01-01

Shielding is provided in a nuclear reactor to absorb the radiations emanating from the core. The energy of these radiations appear in the form of heat. Concrete which is commonly used as a shielding material in nuclear power plants must be able to withstand the temperatures and temperature gradients appearing in the shield due to this heat. High temperatures lead to dehydration of the concrete and in turn reduce the shielding effectiveness of the material. Adequate cooling needs to be provided in these shields in order to limit the maximum temperature. This paper describes a method to estimate steady state and transient temperature distribution in reactor shields. The results due to loss of coolant in the coolant tubes have been studied and presented in the paper. (author). 5 figs

Design of an organic simplified nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Shirvan, Koroush [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States); Forrest, Eric [Primary Standards Laboratory, Sandia National Laboratories, Albuquerque (United States)

2016-08-15

Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

Design of an Organic Simplified Nuclear Reactor

Directory of Open Access Journals (Sweden)

Koroush Shirvan

2016-08-01

Full Text Available Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

Performance Monitoring for Nuclear Safety Related Instrumentation at PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2015-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on performance monitoring for nuclear safety related instrumentation in TRIGA PUSPATI Reactor (RTP) of based on various parameter of reactor safety instrument channel such as log power, linear power, Fuel temperature, coolant temperature will take into consideration. Methodology of performance on estimation and monitoring is to evaluate and analysis of reactor parameters which is important of reactor safety and control. And also to estimate power measurement, differential of log and linear power and fuel temperature during reactor start-up, operation and shutdown .This study also focus on neutron power fluctuation from fission chamber during reactor start-up and operation. This work will present result of performance monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that performance of nuclear safety related instrumentation will improved the reactor control and safety parameter during reactor start-up, operation and shutdown. (author)

Water level monitoring device in nuclear reactor

International Nuclear Information System (INIS)

Miura, Kiyohide; Otake, Tomohiro.

1988-01-01

Purpose: To monitor the water level in a pressure vessel of BWR type nuclear reactors at high accuracy by improving the compensation functions. Constitution: In the conventional water level monitor in a nuclear reactor, if the pressure vessel is displaced by the change of the pressure in the reactor or the temperature of the reactor water, the relative level of the reference water head in a condensation vessel is changed to cause deviation between the actual water level and the indicated water level to reduce the monitoring accuracy. According to the invention, means for detecting the position of the reference water head and means for detection the position in the condensation vessel are disposed to the pressure vessel. Then, relative positional change between the condensation vessel and the reference water head is calculated based on detection sinals from both of the means. The water level is compensated and calculated by water level calculation means based on the relative positional change, water level signals from the level gage and the pressure signals from the pressure gage. As a result, if the pressure vessel is displaced due to the change of the temperature or pressure, it is possible to measure the reactor water level accurately thereby remakably improve the reliability for the water level control in the nuclear reactor. (Horiuchi, T.)

The temperature distribution in a gas core fission reactor

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E.; Dam, H. van; Kuijper, J.C. (Interuniversitair Reactor Inst., Delft (Netherlands)); Kistemaker, J.; Boersma-Klein, W.; Vitalis, F. (FOM-Instituut voor Atoom-en Molecuulfysica, Amsterdam (Netherlands))

1991-01-01

A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters. (author).

The temperature distribution in a gas core fission reactor

International Nuclear Information System (INIS)

Hoogenboom, J.E.; Dam, H. van; Kuijper, J.C.; Kistemaker, J.; Boersma-Klein, W.; Vitalis, F.

1991-01-01

A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters. (author)

Temperature and Doppler coefficients of various space nuclear reactors

International Nuclear Information System (INIS)

Mughabghab, S.F.; Ludewig, H. Schmidt, E.

1993-01-01

Temperature and Doppler feedback effects for a Particle Bed Reactor (PBR) designed to operate as a propulsion reactor are investigated. Several moderator types and compositions fuel enrichments and reactor sizes are considered in this study. From this study it could be concluded that a PBR can be configured which has a negative prompt feedback, zero coolant worth, and a small positive to zero moderator worth. This reactor would put the lowest demands on the control system

Temperature and Doppler Coefficients of Various Space Nuclear Reactors

Science.gov (United States)

Mughabghab, Said F.; Ludewig, Hans; Schmidt, Eldon

1994-07-01

Temperature and Doppler feedback effects for a Particle Bed Reactor (PBR) designed to operate as a propulsion reactor are investigated. Several moderator types and compositions fuel enrichments and reactor sizes are considered in this study. From this study it could be concluded that a PBR can be configured which has a negative prompt feedback, zero coolant worth, and a small positive to zero moderator worth. This reactor would put the lowest demands on the control system.

Nuclear Rocket Engine Reactor

CERN Document Server

Lanin, Anatoly

2013-01-01

The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

Development of methods for monitoring and controlling power in nuclear reactors

International Nuclear Information System (INIS)

Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Santos, Andre Augusto Campagnole dos; Silva, Vitor Vasconcelos Araujo

2012-01-01

Redundancy and diversity are two important criteria for power measurement in nuclear reactors. Other criteria such as accuracy, reliability and response speed are also of major concern. Power monitoring of nuclear reactors is normally done by means of neutronic instruments, i.e. by the measurement of neutron flux. The greater the number of channels for power measuring the greater is the reliability and safety of reactor operations. The aim of this research is to develop new methodologies for on-line monitoring of nuclear reactor power using other reliable processes. One method uses the temperature difference between an instrumented fuel element and the pool water below the reactor core. Another method consists of the steady-state energy balance of the primary and secondary reactor cooling loops. A further method is the calorimetric procedure whereby a constant reactor power is monitored as a function of the temperature-rise rate and the system heat capacity. Another methodology, which does not employ thermal methods, is based on measurement of Cherenkov radiation produced within and around the core. The first three procedures, fuel temperature, energy balance and calorimetric, were implemented in the IPR-R1 TRIGA nuclear research reactor at Belo Horizonte (Brazil) and are the focus of the work described here. Knowledge of the reactor thermal power is very important for precise neutron flux and fuel element burnup calculations. The burnup is linearly dependent on the reactor thermal power and its accuracy is important in the determination of the mass of burned 235 U, fission products, fuel element activity, decay heat power generation and radiotoxicity. The thermal balance method developed in this project is now the standard methodology used for IPR-R1 TRIGA reactor power calibration and the fuel temperature measuring is the most reliable way of on-line monitoring of the reactor power. This research project primarily aims at increasing the reliability and safety of

Separated type nuclear superheating reactor

International Nuclear Information System (INIS)

Hida, Kazuki.

1993-01-01

In a separated type nuclear superheating reactor, fuel assemblies used in a reactor core comprise fuel rods made of nuclear fuel materials and moderator rods made of solid moderating materials such as hydrogenated zirconium. Since the moderating rods are fixed or made detachable, high energy neutrons generated from the fuel rods are moderated by the moderating rods to promote fission reaction of the fuel rods. Saturated steams supplied from the BWR type reactor by the fission energy are converted to high temperature superheated steams while passing through a steam channel disposed between the fuel rods and the moderating rods and supplied to a turbine. Since water is not used but solid moderating materials sealed in a cladding tube are used as moderation materials, isolation between superheated steams and water as moderators is not necessary. Further, since leakage of heat is reduced to improve a heat efficiency, the structure of the reactor core is simplified and fuel exchange is facilitated. (N.H.)

Electrodeless, multi-megawatt reactor for room-temperature, lithium-6/deuterium nuclear reactions

International Nuclear Information System (INIS)

Drexler, J.

1993-01-01

This paper describes a reactor design to facilitate a room-temperature nuclear fusion/fission reaction to generate heat without generating unwanted neutrons, gamma rays, tritium, or other radioactive products. The room-temperature fusion/fission reaction involves the sequential triggering of billions of single-molecule, 6 LiD 'fusion energy pellets' distributed in lattices of a palladium ion accumulator that also acts as a catalyst to produce the molecules of 6 LiD from a solution comprising D 2 O, 6 LiOD with D 2 gas bubbling through it. The D 2 gas is the source of the negative deuterium ions in the 6 LiD molecules. The next step is to trigger a first nuclear fusion/fission reaction of some of the 6 LiD molecules, according to the well-known nuclear reaction: 6 Li + D → 2 4 He + 22.4 MeV. The highly energetic alpha particles ( 4 He nuclei) generated by this nuclear reaction within the palladium will cause shock and vibrations in the palladium lattices, leading to compression of other 6 LiD molecules and thereby triggering a second series of similar fusion/fission reactions, leading to a third series, and so on. The absorption of the kinetic energy in the palladium will, in turn, generate a continuous flow of heat into the heavy water carrier, which would be removed with a heat exchanger. (author)

Assessment of factors that may affect the moisture- and temperature variations in the concrete structures inside nuclear reactor containments

International Nuclear Information System (INIS)

Oxfall, M.; Hassanzadeh, M.; Johansson, P.

2015-01-01

Three factors that may affect the climatic conditions inside Swedish nuclear reactor containments are the outdoor climate, the cooling water temperature and the operational state of the reactor. Those factors that have a considerable affect on the climatic conditions will be identified through comparisons to actual conditions during operation inside reactor containments. Knowledge about the impact of the factors on the climatic conditions is essential when developing a model to determine the prevailing and to predict future conditions in the concrete structures. The comparison between a Pressurized Water Reactor and a Boiling Water Reactor showed that there is no clear similarity, with regard to fluctuation of the indoor climate conditions, between the two reactor types. The indoor climate at a Pressurized Water Reactor seemed not to be affected of changes in the operational state, but it follows the fluctuations of the outdoor temperature and to some extent the water temperature. The Boiling water Reactor did not follow the water or outdoor temperature fluctuations. However, at a full power outage during a short period of time during the operational year the temperature drastically decreased. An operational year of a Boiling water reactor can be divided into three periods, where the first represents the yearly power outage, the second represents the autumn-winter-spring period and the third represents the summer period. The operational year of a pressurized water reactor can't easily be divided into stable periods, since the indoor temperature fluctuation follows the outdoor temperature fluctuation. Based on these measurements a simplified model which includes outer factors is possible for a BWR but more difficult for a PWR. (authors)

Nuclear reactor structural material forming less radioactive corrosion product

International Nuclear Information System (INIS)

Nakazawa, Hiroshi.

1988-01-01

Purpose: To provide nuclear reactor structural materials forming less radioactive corrosion products. Constitution: Ni-based alloys such as inconel alloy 718, 600 or inconel alloy 750 and 690 having excellent corrosion resistance and mechanical property even in coolants at high temperature and high pressure have generally been used as nuclear reactor structural materials. However, even such materials yield corrosion products being attacked by coolants circulating in the nuclear reactor, which produce by neutron irradiation radioactive corrosion products, that are deposited in primary circuit pipeways to constitute exposure sources. The present invention dissolves dissolves this problems by providing less activating nuclear reactor structural materials. That is, taking notice on the fact that Ni-58 contained generally by 68 % in Ni changes into Co-58 under irradiation of neutron thereby causing activation, the surface of nuclear reactor structural materials is applied with Ni plating by using Ni with a reduced content of Ni-58 isotopes. Accordingly, increase in the radiation level of the nuclear reactor structural materials can be inhibited. (K.M.)

Nuclear reactor theory

International Nuclear Information System (INIS)

Sekimoto, Hiroshi

2007-09-01

This textbook is composed of two parts. Part 1 'Elements of Nuclear Reactor Theory' is composed of only elements but the main resource for the lecture of nuclear reactor theory, and should be studied as common knowledge. Much space is therefore devoted to the history of nuclear energy production and to nuclear physics, and the material focuses on the principles of energy production in nuclear reactors. However, considering the heavy workload of students, these subjects are presented concisely, allowing students to read quickly through this textbook. (J.P.N.)

The pebble-bed high-temperature reactor as a source of nuclear process heat. Vol. 10

International Nuclear Information System (INIS)

Schulten, R.; Kugeler, K.; Kugeler, M.; Niessen, H.F.; Roeth-Kamat, M.; Woike, O.

1974-08-01

The necessary development steps, which have to be taken for the construction of a prototype plant for nuclear process heat, are enumerated. In particular, the work which is involved for the development of the nuclear steam-reforming technique, for the further development of the ball-shaped fuel elements at high gas outlet temperatures and for the reactor components, is described in detail. A brief survey of the needs of development of the IHX (intermediate heat exchanger) is given. An attempt is made to give overall time and cost estimates. (orig.) [de

Aspects of nuclear process heat application of very high temperature reactors (VHTR)

International Nuclear Information System (INIS)

Jansing, W.T.; Kugeler, K.

2014-01-01

The different processes of high temperature process application require new concepts for heat exchangers to carry out key process like steam reforming of light hydrocarbons, gasification of coal or biomass, or thermo-chemical cycles for hydrogen production. These components have been tested in the German projects for high temperature development. The intention was always to test at original conditions of temperatures, pressures and gas atmospheres. Furthermore the time of testing should be long as possible, to be able to carry out extrapolations to the real lifetime of components. Partly test times of around 20 000 hours have been reached. Key components, which are discussed in this paper, are: Intermediate heat exchangers to separate the primary reactor side and the secondary process side. Here two components with a power of 10 MW have been tested with the result, that all requirements of a nuclear component with larger power (125 MW) can be fulfilled. The max. primary helium temperature was 950°C, the maximal secondary temperature was 900°C. These were components with helical wounded tubes and U-tubes. In the test facility KVK, which had been built to carry out many special tests on components for helium cycles, furthermore hot gas ducts (with large dimensions), hot gas valves (with large dimensions), steam generators (10 MW), helium circulators, the helium gas purification and special measurements installations for helium cycle have been tested. All these tests delivered a broad know how for the urther development of technologies using helium as working fluid. The total test time of KVK was longer than 20 000 h. In a large test facility for steam reforming (EVAⅡ10 MW, T He =950°C, p He =40 bar, T Reform =800°C) all technical details of the conversion process have been investigated and today the technical feasibility of this process is valuated as given. Two reformer bundles, one with baffles and one with separate guiding tubes for each reformer tube have

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

International Nuclear Information System (INIS)

Vondy, D.R.

1984-07-01

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.

1984-07-01

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

Studies on high temperature research reactor in China

Energy Technology Data Exchange (ETDEWEB)

Xu Yuanhui; Zuo Kanfen [Institute of Nuclear Energy Technology, Tsinghua Univ., Beijing (China)

1999-08-01

China recognises the advantages of Modular HTGRs and has chosen Modular HTGRs as one of advanced reactors to be developed for the further intensive utilisation of nuclear power in the next century. In energy supply systems of the next century, HTGR is supposed to serve: 1. as supplement to water-cooled reactors for electricity generation and 2. as environmentally friendly heat source providing process heat at different temperatures for various applications like heavy oil recovery, coal gasification and liquefaction, etc.. The 10 MW High Temperature Gas-cooled Reactor (HTR-10) is a major project in the energy sector of the Chinese National High Technology Programme as the first step of development of Modular HTGRs in China. Its main objectives are: 1. to acquire know-how in the design, construction and operation of HTGRs, 2. to establish an irradiation and experimental facility, 3. to demonstrate the inherent safety features of Modular HTGR, 4. to test electricity and heat co-generation and closed cycle gas turbine technology and 5. to do research and development work on the nuclear process heat application. Now the HTR-10 is being constructed at the site of Institute of Nuclear Energy Technology (INET). The HTR-10 project is to be carried out in two phases. In the first phase, the reactor with an coolant outlet temperature of 700degC will be coupled with a steam generator providing steam for a steam turbine cycle which works on an electricity and heat co-generation basis. In the second phase, the reactor coolant outlet temperature is planned to be raised to 900degC. As gas turbine cycle and a steam reformer will be coupled to the reactor in addition to the steam turbine cycle. (author)

Studies on high temperature research reactor in China

International Nuclear Information System (INIS)

Xu Yuanhui; Zuo Kanfen

1999-01-01

China recognises the advantages of Modular HTGRs and has chosen Modular HTGRs as one of advanced reactors to be developed for the further intensive utilisation of nuclear power in the next century. In energy supply systems of the next century, HTGR is supposed to serve: 1. as supplement to water-cooled reactors for electricity generation and 2. as environmentally friendly heat source providing process heat at different temperatures for various applications like heavy oil recovery, coal gasification and liquefaction, etc.. The 10 MW High Temperature Gas-cooled Reactor (HTR-10) is a major project in the energy sector of the Chinese National High Technology Programme as the first step of development of Modular HTGRs in China. Its main objectives are: 1. to acquire know-how in the design, construction and operation of HTGRs, 2. to establish an irradiation and experimental facility, 3. to demonstrate the inherent safety features of Modular HTGR, 4. to test electricity and heat co-generation and closed cycle gas turbine technology and 5. to do research and development work on the nuclear process heat application. Now the HTR-10 is being constructed at the site of Institute of Nuclear Energy Technology (INET). The HTR-10 project is to be carried out in two phases. In the first phase, the reactor with an coolant outlet temperature of 700degC will be coupled with a steam generator providing steam for a steam turbine cycle which works on an electricity and heat co-generation basis. In the second phase, the reactor coolant outlet temperature is planned to be raised to 900degC. As gas turbine cycle and a steam reformer will be coupled to the reactor in addition to the steam turbine cycle. (author)

Ultrasonic level and temperature sensor for power reactor applications

International Nuclear Information System (INIS)

Dress, W.B.; Miller, G.N.

1983-01-01

An ultrasonic waveguide employing torsional and extensional acoustic waves has been developed for use as a level and temperature sensor in pressurized and boiling water nuclear power reactors. Features of the device include continuous measurement of level, density, and temperature producing a real-time profile of these parameters along a chosen path through the reactor vessel

9th international conference on high-temperature reactors - coal and nuclear energy for electricity and gas generation

International Nuclear Information System (INIS)

Kelber, G.

1987-01-01

The site of the high-temperatur reactor in the Ruhr region neighbouring on a coal-fired power plant is not accidental. The potential of the high-temperature reactor as a central plant element for the supply of heat for heating purposes and process heat covers also the possibility of coal gasification and liquefaction. Therefore the high-temperature reactor is, in the long term, a ray of hope for the coal region, able to compensate for the production-related competitive disadvantages of local coal. It can contribute to guaranteeing in the long term the task of German hard coal as an essential pillar of our energy supply. The VGB as a technical association of thermal power plant operators is particularly committed to the integration of coal and nuclear energy. Within the bounds of its possibilities, it will contribute to promoting the safe and environmentally beneficial generation of electricity from the two primary energy sources. (orig./DG) [de

Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

International Nuclear Information System (INIS)

Allen, Francis; Bonin, Hugues

2008-01-01

The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU TM nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

Allen, Francis [Director General Nuclear Safety, 280 Slater St, Ottawa, K1A OK2 (Canada); Bonin, Hugues [Royal Military College of Canada, 11 General Crerar Cres, Kingston, K7K 7B4 (Canada)

2008-07-01

The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU{sup TM} nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

Nuclear reactor core safety device

International Nuclear Information System (INIS)

Colgate, S.A.

1977-01-01

The danger of a steam explosion from a nuclear reactor core melt-down can be greatly reduced by adding a gasifying agent to the fuel that releases a large amount of gas at a predetermined pre-melt-down temperature that ruptures the bottom end of the fuel rod and blows the finely divided fuel into a residual coolant bath at the bottom of the reactor. This residual bath should be equipped with a secondary cooling loop

Status of advanced nuclear reactor development in Korea

International Nuclear Information System (INIS)

Kim, H.R.; Kim, K.K.; Kim, Y.W.; Joo, H.K.

2014-01-01

The Korean nuclear industry is facing new challenges to solve the spent fuel storage problem and meet the needs to diversify the application areas of nuclear energy. In order to provide solutions to these challenges, the Korea Atomic Energy Research Institute (KAERI) has been developing advanced nuclear reactors including a Sodium-cooled Fast Reactor, Very High Temperature Gas cooled Reactor (VHTR), and System-integrated Modular Advanced Reactor (SMART) with substantially improved safety, economics, and environment-friendly features. A fast reactor system is one of the most promising options for a reduction of radioactive wastes. The long-term plan for Advanced SFR development in conjunction with the pyro-process was authorized by the Korean Atomic Energy Commission in 2008. The development milestone includes specific design approval of a prototype SFR by 2020, and the construction of a prototype SFR by 2028. KAERI has been carrying out the preliminary design of a 150MWe SFR prototype plant system since 2012. The development of advanced SFR technologies and the basic key technologies necessary for the prototype SFR are also being carried out. By virtue of high-temperature heat, a VHTR has diverse applications including hydrogen production. KAERI launched a nuclear hydrogen project using a VHTR in 2006, which focused on four basic technologies: the development of design tools, very high-temperature experimental technology, TRISO fuel fabrication, and Sulfur-iodine thermo-chemical hydrogen production technology. The technology development project will be continued until 2017. A conceptual reactor design study was started in 2012 as collaboration between industry and government to enhance the early-launching of the nuclear hydrogen development and demonstration (NHDD) project. The goal of the NHDD project is to design and build a nuclear hydrogen demonstration system by 2030. KAERI has developed SMART which is a small-sized advanced integral reactor with a rated

Complete automation of nuclear reactors control

International Nuclear Information System (INIS)

Weill, J.

1955-01-01

The use of nuclear reactor for energy production induces the installation of automatic control systems which need to be safe enough and can adapt to the industrial scale of energy production. These automatic control systems have to insure the constancy of power level and adjust the power produced to the energy demand. Two functioning modes are considered: nuclear plant connected up to other electric production systems as hydraulic or thermic plants or nuclear plants functioning on an independent network. For nuclear plants connected up with other production plants, xenon poisoning and operating cost lead to keep working at maximum power the nuclear reactors. Thus, the power modulation control system will not be considered and only start-up control, safety control, and control systems will be automated. For nuclear power plants working on an independent network, the power modulation control system is needed to economize fuel. It described the automated control system for reactors functioning with constant power: a power measurement system constituted of an ionization chamber and a direct-current amplifier will control the steadfastness of the power produced. For reactors functioning with variable power, the automated power control system will allow to change the power and maintain it steady with all the necessary safety and will control that working conditions under P max and R max (maximum power and maximum reactivity). The effects of temperature and xenon poisoning will also be discussed. Safety systems will be added to stop completely the functioning of the reactor if P max is reached. (M.P.)

Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Khang, Ngo Phu [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

The Dalat nuclear research reactor was reconstructed from TRIGA MARK II reactor, built in 1963 with nominal power of 250 KW, and reached its planned nominal power of 500 kW for the first time in Feb. 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc. (author). 10 refs., 10 figs., 2 tabs.

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

HTGR nuclear power plants: features of the VGR-50 high temperature reactor

International Nuclear Information System (INIS)

Glebov, V.P.; Bogoyavlenskii, R.G.; Glushkov, E.S.; Grebennik, V.N.; Ponomarev-Stepnoi, N.N.; Vinogradov, V.P.

1983-01-01

Current developmental trends in the power industry are guided to an appreciable extent by the increasing shortages of fossil fuels (coal, petroleum, natural gas) and by ecological problems. Assuming a continuing trend in worldwide consumption of energy resources, we see the electric power industry using up 20%, the other 80% (petroleum, coal, natural gas) going into generating industrial process heat and space heat, transportation, the chemical processing industry, the metallurgical industry, and other branches of industry. In the future, nuclear power will have the job of not only meeting the needs of the electric power industry, but also generating process heat. The most promising type of nuclear power plant available for solving complex problems in generation of electric power and heat for technological processes in the metallurgical processing industry and chemical processing industry is the one based around high-temperature reactors

Approximate computation of hydrothermal conditions of nuclear reactor spray ponds

International Nuclear Information System (INIS)

Yarkho, A.A.; Borshchev, V.A.

1990-01-01

An algorithm is presented for determining the evaporation numbers of nuclear reactor spray ponds which provide necessary reactor cooling during its normal operation under given meteorological conditions with account of restrictions on the cooled water temperature at the reactor entrance

Qualitative methods in nuclear reactor dynamics. Issue 23

International Nuclear Information System (INIS)

Goryachenko, V.D.

1983-01-01

Applicability of qualitative methods of the theory of nonlinear oscillations including the bifurcation theory to the problems of nuclear reactor nonlinear dynamics is investigated. Basic statements of the dynamic system qualitative theory on a phase plane and the bifurcation theory of multidimensional dynamic systems are briefly outlined. The model of reactor dynamics with two reactivity temperature coefficients neglecting delayed neutrons, the model of slow process dynamics in a reactor with two reactivity temperature coefficients, the simplified model of reactor dynamics as an object with delay and the model of a reactor with linear feedback are considered. A conclusion is drawn that the usage of the above models allows one to reveal qualitative peculiarities of reactor dynamics creating conditions for more purposeful utilization of more complicated models

Nuclear reactor PBMR and cogeneration

International Nuclear Information System (INIS)

Ramirez S, J. R.; Alonso V, G.

2013-10-01

In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Yoon, Ji Hyun; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-08-15

The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

Completely automated nuclear reactors for long-term operation

International Nuclear Information System (INIS)

Teller, E.; Ishikawa, M.; Wood, L.

1996-01-01

The authors discuss new types of nuclear fission reactors optimized for the generation of high-temperature heat for exceedingly safe, economic, and long-duration electricity production in large, long-lived central power stations. These reactors are quite different in design, implementation and operation from conventional light-water-cooled and -moderated reactors (LWRs) currently in widespread use, which were scaled-up from submarine nuclear propulsion reactors. They feature an inexpensive initial fuel loading which lasts the entire 30-year design life of the power-plant. The reactor contains a core comprised of a nuclear ignitor and a nuclear burn-wave propagating region comprised of natural thorium or uranium, a pressure shell for coolant transport purposes, and automatic emergency heat-dumping means to obviate concerns regarding loss-of-coolant accidents during the plant's operational and post-operational life. These reactors are proposed to be situated in suitable environments at ∼100 meter depths underground, and their operation is completely automatic, with no moving parts and no human access during or after its operational lifetime, in order to avoid both error and misuse. The power plant's heat engine and electrical generator subsystems are located above-ground

A nuclear reactor for district heating

International Nuclear Information System (INIS)

Bancroft, A.R.; Fenton, N.

1989-07-01

Global energy requirements are expected to double over the next 40 years. In the northern hemisphere, many countries consume in excess of 25 percent of their primary energy supply for building heating. Satisfying this need, within the constraints now being acknowledged for sustainable global development, provides an important opportunity for district heating. Fuel-use flexibility, energy and resource conservation, and reduced atmospheric pollution from acid gases and greenhouse gases, are important features offered by district heating systems. Among the major fuel options, only hydro-electricity and nuclear heat completely avoid emissions of combustion gases. To fill the need for an economical nuclear heat source, Atomic Energy of Canada Limited has designed a 10 MW plant that is suitable as a heat source within a network or as the main supply to large individual users. Producing hot water at temperatures below 100 degrees C, it incorporates a small pool-type reactor based on AECL's successful SLOWPOKE Research Reactor. A 2 MW prototype for the commercial unit is now being tested at the Whiteshell Nuclear Research Establishment in Manitoba. With capital costs of $7 million (Canadian), unit energy costs are projected to be $0.02/kWh for a 10 MW unit operating in a heating grid over a 30-year period. By keeping the reactor power low and the water temperature below 100 degrees C, much of the complexity of the large nuclear power plants can be avoided, thus allowing these small, safe nuclear heating systems to be economically viable

SCW Pressure-Channel Nuclear Reactors: Some Design Features and Concepts

International Nuclear Information System (INIS)

Duffey, R.B.; Pioro, I.L.; Gabaraev, B.A.; Kuznetsov, Yu. N.

2006-01-01

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950's and 1960's in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with supercritical water (SCW) became attractive again as the ultimate development path for water-cooling. The main objectives of using SCW in nuclear reactors are 1) to increase the thermal efficiency of modern nuclear power plants (NPPs) from 33 -- 35% to about 40 -- 45%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (∼$ 1000 US/kW). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625 deg. C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia. Design features related to both channels and fuel bundles are discussed in this paper. Also, Russian experience with operating supercritical steam heaters at NPP is presented. The main conclusion is that development of SCW pressure-channel nuclear reactors is feasible and significant benefits can be expected over other thermal energy systems. (authors)

Study on the nuclear heat application system with a high temperature gas-cooled reactor and its safety evaluation (Thesis)

International Nuclear Information System (INIS)

Inaba, Yoshitomo

2008-03-01

Aiming at the realization of the nuclear heat application system with a High Temperature Gas-cooled Reactor (HTGR), research and development on the whole evaluation of the system, the connection technology between the HTGR and a chemical plant such as the safety evaluation against the fire and explosion and the control technology, and the vessel cooling system of the HTGR were carried out. In the whole evaluation of the nuclear heat application system, an ammonia production system using nuclear heat was examined, and the technical subjects caused by the connection of the chemical plant to the HTGR were distilled. After distilling the subjects, the safety evaluation method against the fire and explosion to the reactor, the mitigation technology of thermal disturbance to the reactor, and the reactor core cooling by the vessel cooling system were discussed. These subjects are very important in terms of safety. About the fire and explosion, the safety evaluation method was established by developing the process and the numerical analysis code system. About the mitigation technology of the thermal disturbance, it was demonstrated that the steam generator, which was installed at the downstream of the chemical reactor in the chemical plant, could mitigate the thermal disturbance to the reactor. In order to enhance the safety of the reactor in accidents, the heat transfer characteristic of the passive indirect core cooling system was investigated, and the heat transfer equation considering both thermal radiation and natural convection was developed for the system design. As a result, some technical subjects related to safety in the nuclear heat application system were solved. (author)

Nuclear reaction data and nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Paver, N [University of Trieste (Italy); Herman, M [International Atomic Energy Agency, Vienna (Austria); Gandini, A [ENEA, Rome (Italy)

2001-12-15

These two volumes contain the lecture notes of the workshop 'Nuclear Reaction Data and Nuclear Reactors: Physics, Design and Safety', which was held at the Abdus Salam ICTP in the Spring of 2000. The workshop consisted of five weeks of lecture courses followed by practical computer exercises on nuclear data treatment and design of nuclear power systems. The spectrum of topics is wide enough to timely cover the state-of-the-art and the perspectives of this broad field. The first two weeks were devoted to nuclear reaction models and nuclear data evaluation. Nuclear data processing for applications to reactor calculations was the subject of the third week. On the last two weeks reactor physics and on-going projects in nuclear power generation, waste disposal and safety were presented.

Nuclear reactor container

International Nuclear Information System (INIS)

Ishiyama, Takenori.

1989-01-01

This invention concerns a nuclear reactor container in which heat is removed from a container by external water injection. Heat is removed from the container by immersing the lower portion of the container into water and scattering spary water from above. Thus, the container can be cooled by the spray water falling down along the outer wall of the container to condensate and cool vapors filled in the container upon occurrence of accidents. Further, since the inside of the container can be cooled also during usual operation, it can also serve as a dry well cooler. Accordingly, heat is removed from the reactor container upon occurrence of accidents by the automatic operation of a spray device corresponding to the change of the internal temperature and the pressure in the reactor container. Further, since all of these devices are disposed out of container, maintenance is also facilitated. (I.S.)

Startup method for natural convection type nuclear reactor

International Nuclear Information System (INIS)

Utsuno, Hideaki.

1993-01-01

In a nuclear reactor started by natural convection, no sufficient stability margin can be ensured upon start up. Then, in the present invention, a deaerating operation is conducted before start-up of the reactor, then control rods are withdrawn after conducting the deaerating operation and temperature and pressure are raised by nuclear heating, to obtain a rated power. As a result, reactor power and subcooling at the inlet of the reactor core are within a range of lower than a geysering forming region, thereby enabling to prevent occurence of geysering inherent to the start-up of operation in a natural convection state, shorten the start-up time, as well as remove oxygen dissolved in coolants. (N.H.)

Nuclear reactor instrumentation

International Nuclear Information System (INIS)

Duncombe, E.; McGonigal, G.

1975-01-01

A liquid metal cooled nuclear reactor is described which has an equal number of fuel sub-assemblies and sensing instruments. Each instrument senses temperature and rate of coolant flow of a coolant derived from a group of three sub-assemblies so that an abnormal value for one sub-assembly will be indicated on three instruments thereby providing for redundancy of up to two of the three instruments. The abnormal value may be a precurser to unstable boiling of coolant

The Advanced High-Temperature Reactor (AHTR) for Producing Hydrogen to Manufacture Liquid Fuels

International Nuclear Information System (INIS)

Forsberg, C.W.; Peterson, P.F.; Ott, L.

2004-01-01

Conventional world oil production is expected to peak within a decade. Shortfalls in production of liquid fuels (gasoline, diesel, and jet fuel) from conventional oil sources are expected to be offset by increased production of fuels from heavy oils and tar sands that are primarily located in the Western Hemisphere (Canada, Venezuela, the United States, and Mexico). Simultaneously, there is a renewed interest in liquid fuels from biomass, such as alcohol; but, biomass production requires fertilizer. Massive quantities of hydrogen (H2) are required (1) to convert heavy oils and tar sands to liquid fuels and (2) to produce fertilizer for production of biomass that can be converted to liquid fuels. If these liquid fuels are to be used while simultaneously minimizing greenhouse emissions, nonfossil methods for the production of H2 are required. Nuclear energy can be used to produce H2. The most efficient methods to produce H2 from nuclear energy involve thermochemical cycles in which high-temperature heat (700 to 850 C) and water are converted to H2 and oxygen. The peak nuclear reactor fuel and coolant temperatures must be significantly higher than the chemical process temperatures to transport heat from the reactor core to an intermediate heat transfer loop and from the intermediate heat transfer loop to the chemical plant. The reactor temperatures required for H2 production are at the limits of practical engineering materials. A new high-temperature reactor concept is being developed for H2 and electricity production: the Advanced High-Temperature Reactor (AHTR). The fuel is a graphite-matrix, coated-particle fuel, the same type that is used in modular high-temperature gas-cooled reactors (MHTGRs). The coolant is a clean molten fluoride salt with a boiling point near 1400 C. The use of a liquid coolant, rather than helium, reduces peak reactor fuel and coolant temperatures 100 to 200 C relative to those of a MHTGR. Liquids are better heat transfer fluids than gases

Identification of fast power reactivity effect in nuclear power reactor

International Nuclear Information System (INIS)

Efanov, A.I.; Kaminskas, V.A.; Lavrukhin, V.S.; Rimidis, A.P.; Yanitskene, D.Yu.

1987-01-01

A nuclear power reactor is an object of control with distributed parameters, characteristics of which vary during operation time. At the same time the reactor as the object of control has internal feedback circuits, which are formed as a result of the effects of fuel parameters and a coolant (pressure, temperature, steam content) on the reactor breeding properties. The problem of internal feedback circuit identification in a nuclear power reactor is considered. Conditions for a point reactor identification are obtained and algorithms of parametric identification are constructed. Examples of identification of fast power reactivity effect for the RBMK-1000 reactor are given. Results of experimental testing have shown that the developed method of fast power reactivity effect identification permits according to the data of normal operation to construct adaptive models for the point nuclear reactor, designed for its behaviour prediction in stationary and transition operational conditions. Therefore, the models considered can be used for creating control systems of nuclear power reactor thermal capacity (of RBMK type reactor, in particular) which can be adapted to the change in the internal feedback circuit characteristics

The future of the low temperature district heating reactor

International Nuclear Information System (INIS)

Lu Yingzhong; Wang Dazhong; Ma Changwen; Dong Duo; Tian Jiafu.

1984-01-01

In this paper, the role, development and situation of the low temperature district heating reactor (LTDHR) are briefly summarized. There are four types of LTDHR. They are PWR, reactor with boiling in the chimney, organic reactor and swimming pool reactor. The features of these reactors are introduced. The situation and role of the LTDHR in the future of the energy system are also discussed. The experiment on nuclear district heating with the swimming pool reactor in Qinghua Univ. is described briefly. (Author)

Study of reactivity of fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Rammsy, J.E.M.

1985-01-01

The reactor physics calculations of a 19 module Fluidized Bed Nuclear Reactor using Leopard and Odog codes are performed. The behaviour of the reactor was studied by calculating the reactivity of the reactor as a function of the parameters governing the operational and accidental conditions of the reactor. The effects of temperature, pressure, and vapor generation in the core on the reactivity are calculated. Also the start up behaviour of the reactor is analyzed. For the purpose of the study of a prototype research reactor, the calculations on a one module reactor have been performed. (Author) [pt

Dynamic Complexity Study of Nuclear Reactor and Process Heat Application Integration

International Nuclear Information System (INIS)

Taylor, J'Tia Patrice; Shropshire, David E.

2009-01-01

This paper describes the key obstacles and challenges facing the integration of nuclear reactors with process heat applications as they relate to dynamic issues. The paper also presents capabilities of current modeling and analysis tools available to investigate these issues. A pragmatic approach to an analysis is developed with the ultimate objective of improving the viability of nuclear energy as a heat source for process industries. The extension of nuclear energy to process heat industries would improve energy security and aid in reduction of carbon emissions by reducing demands for foreign derived fossil fuels. The paper begins with an overview of nuclear reactors and process application for potential use in an integrated system. Reactors are evaluated against specific characteristics that determine their compatibility with process applications such as heat outlet temperature. The reactor system categories include light water, heavy water, small to medium, near term high-temperature, and far term high temperature reactors. Low temperature process systems include desalination, district heating, and tar sands and shale oil recovery. High temperature processes that support hydrogen production include steam reforming, steam cracking, hydrogen production by electrolysis, and far-term applications such as the sulfur iodine chemical process and high-temperature electrolysis. A simple static matching between complementary systems is performed; however, to gain a true appreciation for system integration complexity, time dependent dynamic analysis is required. The paper identifies critical issues arising from dynamic complexity associated with integration of systems. Operational issues include scheduling conflicts and resource allocation for heat and electricity. Additionally, economic and safety considerations that could impact the successful integration of these systems are considered. Economic issues include the cost differential arising due to an integrated system

Assessment of mass fraction and melting temperature for the application of limestone concrete and siliceous concrete to nuclear reactor basemat considering molten core-concrete interaction

Energy Technology Data Exchange (ETDEWEB)

Lee, Ho Jae; Kim, Do Gyeum [Korea Institute of Civil Engineering and Building Technology, Goyang (Korea, Republic of); Cho, Jae Leon [Korea Hydro and Nuclear Power Co., Ulsan (Korea, Republic of); Yoon, Eui Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Cho, Myung Suk [Korea Hydro and Nuclear Power Co., Central Research Institute, Daejeon (Korea, Republic of)

2016-04-15

Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies the mass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The H2O content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of CO2 necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core-concrete interaction analysis.

Potentialities of high temperature reactors (HTR)

International Nuclear Information System (INIS)

Hittner, D.

2001-01-01

This articles reviews the assets of high temperature reactors concerning the amount of radioactive wastes produced. 2 factors favors HTR-type reactors: high thermal efficiency and high burn-ups. The high thermal efficiency is due to the high temperature of the coolant, in the case of the GT-MHR project (a cooperation between General Atomic, Minatom, Framatome, and Fuji Electric) designed to burn Russian military plutonium, the expected yield will be 47% with an outlet helium temperature of 850 Celsius degrees. The high temperature of the coolant favors a lot of uses of the heat generated by the reactor: urban heating, chemical processes, or desalination of sea water.The use of a HTR-type reactor in a co-generating way can value up to 90% of the energy produced. The high burn-up is due to the technology of HTR-type fuel that is based on encapsulation of fuel balls with heat-resisting materials. The nuclear fuel of Fort-Saint-Vrain unit (Usa) has reached values of burn-ups from 100.000 to 120.000 MWj/t. It is shown that the quantity of unloaded spent fuel can be divided by 4 for the same amount of electricity produced, in the case of the GT-MHR project in comparison with a light water reactor. (A.C.)

System Evaluation and Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen-Production Plant

International Nuclear Information System (INIS)

Harvego, E.A.; McKellar, M.G.; Sohal, M.S.; O'Brien, J.E.; Herring, J.S.

2010-01-01

A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540 C and 900 C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating current (AC) to direct current (DC) conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.1% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

Solution of heat removal from nuclear reactors by natural convection

Directory of Open Access Journals (Sweden)

Zitek Pavel

2014-03-01

Full Text Available This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR.The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

Assessment of very high temperature reactors in process applications

International Nuclear Information System (INIS)

Jones, J.E. Jr.; Spiewak, I.; Gambill, W.R.

1976-01-01

In April 1974, the United States Energy Research and Development Administration (ERDA) authorized General Atomic Company, General Electric Company, and Westinghouse Astronuclear Laboratory to assess the available technology for producing process heat utilizing a very high temperature nuclear reactor (VHTR). The VHTR is defined as a gas-cooled graphite-moderated reactor. Oak Ridge National Laboratory has been given a lead role in evaluating the VHTR reactor studies and potential applications of the VHTR. Process temperatures up to the 760 to 871 0 C range appear to be achievable with near-term technology. The major development considerations are high temperature materials, the safety questions (especially regarding the need for an intermediate heat exchanger) and the process heat exchanger. The potential advantages of the VHTR over competing fossil energy sources are conservation of fossil fuels and reduced atmospheric impacts. Costs are developed for nuclear process heat supplied from a 3000-MW(th) VHTR. The range of cost in process applications is competitive with current fossil fuel alternatives

Nuclear reactor neutron shielding

Science.gov (United States)

Speaker, Daniel P; Neeley, Gary W; Inman, James B

2017-09-12

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

Design Configurations and Coupling High Temperature Gas-Cooled Reactor and Hydrogen Plant

International Nuclear Information System (INIS)

Chang H. Oh; Eung Soo Kim; Steven Sherman

2008-01-01

The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood

Evaluation of two processes of hydrogen production starting from energy generated by high temperature nuclear reactors; Evaluacion de dos procesos de produccion de hidrogeno a partir de energia generada por reactores nucleares de alta temperatura

Energy Technology Data Exchange (ETDEWEB)

Valle H, J., E-mail: jvalle@upmh.edu.mx [Universidad Politecnica Metropolitana de Hidalgo, Boulevard Acceso a Tolcayuca 1009, Ex-Hacienda San Javier, 43860 Tolcayuca, Hidalgo (Mexico)

2013-10-15

In this work an evaluation to two processes of hydrogen production using energy generated starting from high temperature nuclear reactors (HTGR's) was realized. The evaluated processes are the electrolysis of high temperature and the thermo-chemistry cycle Iodine-Sulfur. The electrolysis of high temperature, contrary to the conventional electrolysis, allows reaching efficiencies of up to 60% because when increasing the temperature of the water, giving thermal energy, diminishes the electric power demand required to separate the molecule of the water. However, to obtain these efficiencies is necessary to have water vapor overheated to more than 850 grades C, temperatures that can be reached by the HTGR. On the other hand the thermo-chemistry cycle Iodine-Sulfur, developed by General Atomics in the 1970 decade, requires two thermal levels basically, the great of them to 850 grades C for decomposition of H{sub 2}SO{sub 4} and another minor to 360 grades C approximately for decomposition of H I, a high temperature nuclear reactor can give the thermal energy required for the process whose products would be only hydrogen and oxygen. In this work these two processes are described, complete models are developed and analyzed thermodynamically that allow to couple each hydrogen generation process to a reactor HTGR that will be implemented later on for their dynamic simulation. The obtained results are presented in form of comparative data table of each process, and with them the obtained net efficiencies. (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.)

Nuclear data and reactor physics activities in Indonesia

Energy Technology Data Exchange (ETDEWEB)

Liem, P.H. [National Atomic Energy Agency, Tangerang (Indonesia). Center for Multipurpose Reactor

1998-03-01

The nuclear data and reactor physics activities in Indonesia, especially, in the National Atomic Energy Agency are presented. In the nuclear data field, the Agency is now taking the position of a user of the main nuclear data libraries such as JENDL and ENDF/B. These nuclear data libraries become the main sources for producing problem dependent cross section sets that are needed by cell calculation codes or transport codes for design, analysis and safety evaluation of research reactors. In the reactor physics field, besides utilising the existing core analysis codes obtained from bilateral and international co-operation, the Agency is putting much effort to self-develop Batan`s codes for reactor physics calculations, in particular, for research reactor and high temperature reactor design, analysis and fuel management. Under the collaboration with JAERI, Monte Carlo criticality calculations on the first criticality of RSG GAS (MPR-30) first core were done using JAERI continuous energy, vectorized Monte Carlo code, MVP, with JENDL-3.1 and JENDL-3.2 nuclear data libraries. The results were then compared with the experiment data collected during the commissioning phase. Monte Carlo calculations with both JENDL-3.1 and -3.2 libraries produced k{sub eff} values with excellent agreement with experiment data, however, systematically, JENDL-3.2 library showed slightly higher k{sub eff} values than JENDL-3.1 library. (author)

Nuclear-Mechanical Coupling: Small Amplitude Mechanical Vibrations and High Amplitude Power Oscillations in Nuclear Reactors

International Nuclear Information System (INIS)

Suarez Antola, R.

2008-11-01

The cores of nuclear reactors, including its structural parts and cooling fluids, are complex mechanical systems able to vibrate in a set of normal modes and frequencies, if suitable perturbed. The cyclic variations in the strain state of the core materials may produce changes in density. Changes in density modify the reactivity. Changes in reactivity modify thermal power. Modifications in thermal power produce variations in temperature fields. Variations in temperature produce variations in strain due to thermal-elastic effects. If the variation of the temperature field is fast enough and if the Doppler Effect and other stabilizing prompt effects in the fuel are weak enough, a fast oscillatory instability could be produced, coupled with mechanical vibrations of small amplitude. A recently constructed, simple mathematical model of nuclear reactor kinetics, that improves the one due to A.S. Thompson, is reviewed. It was constructed in order to study, in a first approximation, the stability of the reactor: a nonlinear nuclear-thermal oscillator (that corresponds to reactor point kinetics with thermal-elastic feedback and with frozen delayed neutron effects) is coupled nonlinearly with a linear mechanical-thermal oscillator (that corresponds to the first normal mode of mechanical vibrations excited by thermo-elastic effects). This mathematical model is studied here from the standpoint of mechanical vibrations. It is shown how, under certain conditions, a suitable mechanical perturbation could elicit fast and growing oscillatory instabilities in the reactor power. Applying the asymptotic method due to Krylov, Bogoliubov and Mitropolsky, analytical formulae that may be used in the calculation of the time varying amplitude and phase of the mechanical oscillations are given, as functions of the mechanical, thermal and nuclear parameters of the reactor. The consequences for the mechanical integrity of the reactor are assessed. Some conditions, mainly, but not exclusively

Fast Transient And Spatially Non-Homogenous Accident Analysis Of Two-Dimensional Cylindrical Nuclear Reactor

International Nuclear Information System (INIS)

Yulianti, Yanti; Su'ud, Zaki; Waris, Abdul; Khotimah, S. N.; Shafii, M. Ali

2010-01-01

The research about fast transient and spatially non-homogenous nuclear reactor accident analysis of two-dimensional nuclear reactor has been done. This research is about prediction of reactor behavior is during accident. In the present study, space-time diffusion equation is solved by using direct methods which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference discretization method is solved by using iterative methods ADI (Alternating Direct Implicit). The indication of accident is decreasing macroscopic absorption cross-section that results large external reactivity. The power reactor has a peak value before reactor has new balance condition. Changing of temperature reactor produce a negative Doppler feedback reactivity. The reactivity will reduce excess positive reactivity. Temperature reactor during accident is still in below fuel melting point which is in secure condition.

Nuclear reactor coolant channels

International Nuclear Information System (INIS)

Macbeth, R.V.

1978-01-01

A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

Nuclear heat source component design considerations for HTGR process heat reactor plant concept

International Nuclear Information System (INIS)

McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

1982-05-01

The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported [via an intermediate heat exchanger (IHX)] to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

A new fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Sefidvash, F.

1986-01-01

A new nuclear reactor design based on the fluidized bed concept is proposed. A current design utilizes spherical fuel of slightly enriched Zircaloy-clad uranium dioxide fluidized by light water under pressure. The reactor is modular in system; therefore, any size reactor can be constructed from the basic standard modul. The reactor physics calculations show that reactivity increases with porosity to a maximum value and thereafter decreases. This produces inherent safety and eliminates the need for control rods and burnable poisons. The heat transfer calculations show that the maximum power extracted from the reactor core is not limited to the material temperature limits but to the maximum mass flow of coolant, which corresponds to the desired operating porosity. Design simplicity and inherent safety make it an attractive small reactor design. (Author) [pt

Numerical modeling of a nuclear production reactor cooling lake

International Nuclear Information System (INIS)

Hamm, L.L.; Pepper, D.W.

1987-01-01

A finite element model has been developed which predicts flow and temperature distributions within a nuclear reactor cooling lake at the Savannah River Plant near Aiken, South Carolina. Numerical results agree with values obtained from a 3-D EPA numerical lake model and actual measurements obtained from the lake. Because the effluent water from the reactor heat exchangers discharges directly into the lake, downstream temperatures at mid-lake could exceed the South Carolina DHEC guidelines for thermal exchanges during the summer months. Therefore, reactor power was reduced to maintain temperature compliance at mid-lake. Thermal mitigation measures were studied that included placing a 6.1 m deep fabric curtain across mid-lake and moving the reactor outfall upstream. These measurements were calculated to permit about an 8% improvement in reactor power during summer operation

Nuclear graphite for high temperature reactors

International Nuclear Information System (INIS)

Marsden, B.J.

2001-01-01

The cores and reflectors in modern High Temperature Gas Cooled Reactors (HTRs) are constructed from graphite components. There are two main designs; the Pebble Bed design and the Prism design. In both of these designs the graphite not only acts as a moderator, but is also a major structural component that may provide channels for the fuel and coolant gas, channels for control and safety shut off devices and provide thermal and neutron shielding. In addition, graphite components may act as a heat sink or conduction path during reactor trips and transients. During reactor operation, many of the graphite component physical properties are significantly changed by irradiation. These changes lead to the generation of significant internal shrinkage stresses and thermal shut down stresses that could lead to component failure. In addition, if the graphite is irradiated to a very high irradiation dose, irradiation swelling can lead to a rapid reduction in modulus and strength, making the component friable.The irradiation behaviour of graphite is strongly dependent on its virgin microstructure, which is determined by the manufacturing route. Nevertheless, there are available, irradiation data on many obsolete graphites of known microstructures. There is also a well-developed physical understanding of the process of irradiation damage in graphite. This paper proposes a specification for graphite suitable for modern HTRs. (author)

Issues of high-burnup fuel for advanced nuclear reactors

International Nuclear Information System (INIS)

Belac, J.; Milisdoerfer, L.

2004-12-01

A brief description is given of nuclear fuels for Generation III+ and IV reactors, and the major steps needed for a successful implementation of new fuels in prospective types of newly designed power reactors are outlined. The following reactor types are discussed: gas cooled fast reactors, heavy metal (lead) cooled fast reactors, molten salt cooled reactors, sodium cooled fast reactors, supercritical water cooled reactors, and very high temperature reactors. The following are regarded as priority areas for future investigations: (i) spent fuel radiotoxicity; (ii) proliferation volatility; (iii) neutron physics characteristics and inherent safety element assessment; technical and economic analysis of the manufacture of advanced fuels; technical and economic analysis of the fuel cycle back end, possibilities of spent nuclear fuel reprocessing, storage and disposal. In parallel, work should be done on the validation and verification of analytical tools using existing and/or newly acquired experimental data. (P.A.)

Modification of reference temperature program in reactor regulating system

Energy Technology Data Exchange (ETDEWEB)

Yu, Sung Sik; Lee, Byung Jin; Kim, Se Chang; Cheong, Jong Sik [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Kim, Ji In; Doo, Jin Yong [Korea Electric Power Cooperation, Yonggwang (Korea, Republic of)

1999-12-31

In Yonggwang nuclear units 3 and 4 currently under commercial operation, the cold temperature was very close to the technical specification limit of 298 deg C during initial startup testing, which was caused by the higher-than-expected reactor coolant system flow. Accordingly, the reference temperature (Tref) program needed to be revised to allow more flexibility for plant operations. In this study, the method of a specific test performed at Yonggwang nuclear unit 4 to revise the Tref program was described and the test results were discussed. In addition, the modified Tref program was evaluated on its potential impacts on system performance and safety. The methods of changing the Tref program and the associated pressurizer level setpoint program were also explained. Finally, for Ulchin nuclear unit 3 and 4 currently under initial startup testing, the effects of reactor coolant system flow rate on the coolant temperature were evaluated from the thermal hydraulic standpoint and an optimum Tref program was recommended. 6 refs., 4 figs., 2 tabs. (Author)

Modification of reference temperature program in reactor regulating system

Energy Technology Data Exchange (ETDEWEB)

Yu, Sung Sik; Lee, Byung Jin; Kim, Se Chang; Cheong, Jong Sik [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Kim, Ji In; Doo, Jin Yong [Korea Electric Power Cooperation, Yonggwang (Korea, Republic of)

1998-12-31

In Yonggwang nuclear units 3 and 4 currently under commercial operation, the cold temperature was very close to the technical specification limit of 298 deg C during initial startup testing, which was caused by the higher-than-expected reactor coolant system flow. Accordingly, the reference temperature (Tref) program needed to be revised to allow more flexibility for plant operations. In this study, the method of a specific test performed at Yonggwang nuclear unit 4 to revise the Tref program was described and the test results were discussed. In addition, the modified Tref program was evaluated on its potential impacts on system performance and safety. The methods of changing the Tref program and the associated pressurizer level setpoint program were also explained. Finally, for Ulchin nuclear unit 3 and 4 currently under initial startup testing, the effects of reactor coolant system flow rate on the coolant temperature were evaluated from the thermal hydraulic standpoint and an optimum Tref program was recommended. 6 refs., 4 figs., 2 tabs. (Author)

New ceramics for nuclear industry. Case of fission and fusion reactors

International Nuclear Information System (INIS)

Yvars, M.

1979-10-01

The ceramics used in the nuclear field are described as is their behaviour under radiation. 1) Power reactors - nuclear fission. Ceramics enter into the fabrication of nuclear fuels: oxides, carbides, uranium or plutonium nitrides or oxy-nitrides. Silicon carbide SiC is used for preparing the fuels of helium cooled high temperature reactors. Its use is foreseen in the design of gas high temperature gas thermal exchangers, as is silicon nitride (Si 3 N 4 ). In the materials for safety or control rods, the intense neutron flows induce nuclear reactions which increase the temperature of the neutron absorbing material. Boron carbide B 4 C, rare earth oxides Ln 2 O 3 , or B 4 C-Cu or B 4 C-Al cermets are employed. Burnable poison materials are formed of Al 2 O 3 -B 4 C or Al 2 O 3 -Ln 2 O 3 cermets. The moderators of thermal neutron reactors are in high purety polycrystalline graphite. For the thermal insulation of reactor vessels and jackets, honeycomb ceramics are used as well as ceramic fibres on an increasing scale (kaolin, alumina and other fibres). 2) fusion reactors (Tokomak). These require refractory materials with a low atomic number. Carbon fibres, boron carbide, some borons (Al B 12 ), silicon nitrides and oxy-nitrides and high density alumina are the substances considered [fr

Physics of nuclear reactors

International Nuclear Information System (INIS)

Baeten, Peter

2006-01-01

This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

High temperature fast reactor for hydrogen production in Brazil; Reator nuclear rapido de altissima temperatura para producao de hidrogenio no Brasil

Energy Technology Data Exchange (ETDEWEB)

Nascimento, Jamil A. do; Ono, Shizuca; Guimaraes, Lamartine N.F. [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados]. E-mail: jamil@ieav.cta.br

2008-07-01

The main nuclear reactors technology for the Generation IV, on development phase for utilization after 2030, is the fast reactor type with high temperature output to improve the efficiency of the thermo-electric conversion process and to enable applications of the generated heat in industrial process. Currently, water electrolysis and thermo chemical cycles using very high temperature are studied for large scale and long-term hydrogen production, in the future. With the possible oil scarcity and price rise, and the global warming, this application can play an important role in the changes of the world energy matrix. In this context, it is proposed a fast reactor with very high output temperature, {approx} 1000 deg C. This reactor will have a closed fuel cycle; it will be cooled by lead and loaded with nitride fuel. This reactor may be used for hydrogen, heat and electricity production in Brazil. It is discussed a development strategy of the necessary technologies and some important problems are commented. The proposed concept presents characteristics that meet the requirements of the Generation IV reactor class. (author)

Optical Fibers in Nuclear Reactor Radiation Environments.

Science.gov (United States)

Holcomb, David Eugene

1992-01-01

A performance evaluation of fiber optics under radiation conditions similar to those encountered in nuclear power plants is reported. The evaluation was accomplished by the creation of an analytical model for atomic scale radiation damage in silica glass and by the execution of an extensive fiber performance measurement program. The analytic model calculates displacement and electronic damage rates for silica glass subjected to a specified nuclear reactor radiation environment. It accomplishes this by first generating the primary charged particle spectrum produced in silica irradiated in a nuclear reactor. The resultant spectra are then applied to the integral equations describing radiation damage in polyatomic solids. The experimental measurements were selected to span the range of fiber types, radiation environments, temperatures, and light powers expected to be used in nuclear power plants. The basic experimental protocol was to expose the optical fibers to either a nuclear reactor or a ^{60}Co radiation environment while simultaneously monitoring fiber light transmission. Experimental temperatures were either ~23 ^circC or ~100 ^circC and light powers were either -30 dBm or -60 dBm. Measurements were made at each of the three standard communications wavelengths (850 nm, 1300 nm, and 1550 nm). Several conclusions are made based on these performance measurements. First, even near the core of a nuclear reactor the vast majority of the dose to silica glass is due to gamma rays. Even with the much lower doses (factor of roughly 40) neutrons cause much more displacement damage than gamma rays (35 times the oxygen displacement rate and 500 times the silicon displacement rate). Even with neutrons having many times the displacement rate as compared with gamma rays, little if any difference is observed in the transmission losses for gamma only as compared to mixed neutron/gamma transmission losses. Therefore, atomic displacement is not a significant damage mechanism for

Coupling of high temperature nuclear reactor with chemical plant by means of steam loop with heat pump

Directory of Open Access Journals (Sweden)

Kopeć Mariusz

2017-01-01

Full Text Available High temperature nuclear reactors (HTR can be used as an excellent, emission-free source of technological heat for various industrial applications. Their outlet helium temperature (700°-900°C allows not only for heat supply to all processes below 600°C (referred to as “steam class”, but also enables development of clean nuclear-assisted hydrogen production or coal liquefaction technologies with required temperatures up to 900°C (referred to as “chemical class”. This paper presents the results of analyses done for various configurations of the steam transport loop coupled with the high-temperature heat pump designed for “chemical class” applications. The advantages and disadvantages as well as the key issues are discussed in comparison with alternative solutions, trying to answer the question whether the system with the steam loop and the hightemperature heat pump is viable and economically justified.

Micro-structured nuclear fuel and novel nuclear reactor concepts for advanced power production

International Nuclear Information System (INIS)

Popa-Simil, Liviu

2008-01-01

Many applications (e.g. terrestrial and space electric power production, naval, underwater and railroad propulsion and auxiliary power for isolated regions) require a compact-high-power electricity source. The development of such a reactor structure necessitates a deeper understanding of fission energy transport and materials behavior in radiation dominated structures. One solution to reduce the greenhouse-gas emissions and delay the catastrophic events' occurrences may be the development of massive nuclear power. The actual basic conceptions in nuclear reactors are at the base of the bottleneck in enhancements. The current nuclear reactors look like high security prisons applied to fission products. The micro-bead heterogeneous fuel mesh gives the fission products the possibility to acquire stable conditions outside the hot zones without spilling, in exchange for advantages - possibility of enhancing the nuclear technology for power production. There is a possibility to accommodate the materials and structures with the phenomenon of interest, the high temperature fission products free fuel with near perfect burning. This feature is important to the future of nuclear power development in order to avoid the nuclear fuel peak, and high price increase due to the immobilization of the fuel in the waste fuel nuclear reactor pools. (author)

Abnormality diagnosis device for nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Utsunomiya, Kazuhiro; Oyama, Shinmi; Sakaba, Hideo

1989-02-21

According to the present invention, abnormality such as abnormal increase of temperature in a nuclear reactor is detected to send a signal to control rod drives, etc. thereby stopping the operation of the nuclear reactor. Receiving/transmission device transmits a signal for conducting normal operation of an abnormality information section, as well as receives an echo signal from the abnormality information section to transmit an abnormal signal to a reactor protection system. The abnormality information section is disposed to fuel assemblies, receives a signal from the receiving/transmission device for conducting the normal operation to transmit a normal echo signal, as well as changes the echo signal when detecting the nuclear reactor abnormality. By the foregoing method, since the abnormality information section is disposed to the fuel assemblies, various effects can be attained such as: (1) there is no response delay from the occurrence of abnormality to emergency counter measure after detection, (2) high burnup degree for fuels can thus be possible to improve the economical property, (3) the abnormality information section can be taken out from the reactor container together with fuel assemablies by an existent take-out mechanism and (4) since wireless transmission and reception are established between the receiving/transmission device and the abnormality information section, cables are not required in the container. (K.M.).

Control room conceptual design of nuclear power plant with multiple modular high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Jia Qianqian; Qu Ronghong; Zhang Liangju

2014-01-01

A conceptual design of the control room layout for the nuclear power plant with multiple modular high temperature gas-cooled reactors has been developed. The modular high temperature gas-cooled reactors may need to be grouped to produce as much energy as a utility demands to realize the economic efficiency. There are many differences between the multi-modular plant and the current NPPs in the control room. These differences may include the staffing level, the human-machine interface design, the operation mode, etc. The potential challenges of the human factor engineering (HFE) in the control room of the multi-modular plant are analyzed, including the operation workload of the multi-modular tasks, how to help the crew to keep situation awareness of all modules, and how to support team work, the control of shared system between modules, etc. A concept design of control room for the multi-modular plant is presented based on the design aspect of HTR-PM (High temperature gas-cooled reactor pebble bed module). HFE issues are considered in the conceptual design of control room for the multi-modular plant and some design strategies are presented. As a novel conceptual design, verifications and validations are needed, and focus of further work is sketch out. (author)

Perspective of nuclear energy and advanced reactors

International Nuclear Information System (INIS)

Lopez Jimenez, J.; Cobian, J.

2007-01-01

Future nuclear energy growth will be the result of the contributions of every single plant being constructed or projected at present as it is connected to the grid. As per IAEA, there exists presently 34 nuclear power plants under construction 81 with the necessary permits and funding and 223 proposed, which are plants seriously pursuing permits and financing. This means that in a few decades the number of nuclear power plants in operation will have doubled. This growth rate is characterised by the incorporation of new countries to the nuclear club and the gradually increasing importance of Asian countries. During this expansive phase, generation III and III+designs are or will be used. These designs incorporate the experience from operating plants, and introduce innovations on rationalization design efficiency and safety, with emphasis on passive safety features. In a posterior phase, generation IV designs, presently under development, will be employed. Generation IV consists of several types of reactors (fast reactors, very high temperature reactors, etc), which will improve further sustain ability, economy, safety and reliability concepts. The described situation seems to lead to a renaissance of the nuclear energy to levels hardly thinkable several years ago. (Author)

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.)

Low-temperature thermionics in space nuclear power systems with the safe-type fast reactor

International Nuclear Information System (INIS)

Zrodnikov, A.V.; Yarygin, V.I.; Lazarenko, G.E.; Zabudko, A.N.; Ovcharenko, M.K.; Pyshko, A.P.; Mironov, V.S.; Kuznetsov, R.V.

2007-01-01

The potentialities of the use of the low-temperature thermionic converters (TIC) with the emitter temperature ≤ 1500 K in the space nuclear power system (SNPS) with the SAFE-type (Safe Affordable Fission Engine) fast reactor proposed and developed by common efforts of American experts have been considered. The main directions of the 'SAFE-300-TEG' SNPS (300 kW(thermal)) design update by replacing the thermoelectric converters with the low-temperature high-performance thermionic converters (with the barrier index V B ≤ 1.9 eV and efficiency ≥ 10%) meant for a long-term operation (5 years at least) as the components of the SAFE-300-TIC SNPS for a Lunar base have been discussed. The concept of the SNPS with the SAFE-type fast reactor and low-temperature TICs with specific electric power of about 1.45 W/cm 2 as the components of the SAFE-300-TIC system meeting the Nasa's initial requirements to a Lunar base with the electric power demand of about 30 kW(electrical) for robotic mission has been considered. The results, involving optimization and mass-and-size estimation, show that the SAFE-300-TIC system meets the initial requirements by Nasa to the lunar base power supply. The main directions of the system update aimed at the output electric power increase up to 100 kW(electrical) have also been presented. (authors)

Assessment of Mass Fraction and Melting Temperature for the Application of Limestone Concrete and Siliceous Concrete to Nuclear Reactor Basemat Considering Molten Core–Concrete Interaction

Directory of Open Access Journals (Sweden)

Hojae Lee

2016-04-01

Full Text Available Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies the mass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The H2O content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of CO2 necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core–concrete interaction analysis.

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

Radiation embrittlement of Spanish nuclear reactor pressure vessel steels

International Nuclear Information System (INIS)

Bros, J.; Ballesteros, A.; Lopez, A.

1993-01-01

Commercial pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear power plants contain a series of pressure vessel steel surveillance capsules as the principal means of monitoring radiation effects on the pressure vessel. Changes in fracture toughness are more severe in surveillance capsules than in reactor vessel materials because of their proximity of the reactor core. Therefore, it is possible to predict changes in fracture toughness of the reactor vessel materials. This paper describes the status of the reactor vessel surveillance program relating to Spanish nuclear power plants. To date, twelve capsules have been removed and analyzed from seven of the nine Spanish reactors in operation. The results obtained from the analysis of these capsules are compared with the predictions of the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2, by means of measured and expected increase of the nil-ductility transition reference temperature (RT NDT ). The comparison is made considering the different variables normally included in the studies of radiation response of reactor pressure vessel materials, such as copper content of steel, level of neutron fluence above 1 MeV, base metal or weld metal, and so forth. The surveillance data have been used for determining the adjusted reference temperatures and upper shelf energies at any time. The results have shown that the seven pressure vessels are in excellent condition to continue operating with safety against brittle fracture beyond the design life, without the need to recuperate the degraded properties of the materials by annealing of the vessel

Nuclear reactor buildings

International Nuclear Information System (INIS)

Nagashima, Shoji; Kato, Ryoichi.

1985-01-01

Purpose: To reduce the cost of reactor buildings and satisfy the severe seismic demands in tank type FBR type reactors. Constitution: In usual nuclear reactor buildings of a flat bottom embedding structure, the flat bottom is entirely embedded into the rock below the soils down to the deck level of the nuclear reactor. As a result, although the weight of the seismic structure can be decreased, the amount of excavating the cavity is significantly increased to inevitably increase the plant construction cost. Cross-like intersecting foundation mats are embedded to the building rock into a thickness capable withstanding to earthquakes while maintaining the arrangement of equipments around the reactor core in the nuclear buildings required by the system design, such as vertical relationship between the equipments, fuel exchange systems and sponteneous drainings. Since the rock is hard and less deformable, the rigidity of the walls and the support structures of the reactor buildings can be increased by the embedding into the rock substrate and floor responsivity can be reduced. This enables to reduce the cost and increasing the seismic proofness. (Kamimura, M.)

Nuclear Research Center IRT reactor dynamics calculation

International Nuclear Information System (INIS)

Aleman Fernandez, J.R.

1990-01-01

The main features of the code DIRT, for dynamical calculations are described in the paper. With the results obtained by the program, an analysis of the dynamic behaviour of the Research Reactor IRT of the Nuclear Research Center (CIN) is performed. Different transitories were considered such as variation of the system reactivity, coolant inlet temperature variation and also variations of the coolant velocity through the reactor core. 3 refs

Research on reactor physics using the Very High Temperature Reactor Critical Assembly (VHTRC)

International Nuclear Information System (INIS)

Akino, Fujiyoshi

1988-01-01

The High Temperature Engineering Test Reactor (HTTR), of which the research and development are advanced by Japan Atomic Energy Research Institute, is planned to apply for the permission of installation in fiscal year 1988, and to start the construction in the latter half of fisical year 1989. As the duty of reactor physics research, the accuracy of the nuclear data is to be confirmed, the validity of the nuclear design techniques is to be inspected, and the nuclear safety of the HTTR core design is to be verified. Therefore, by using the VHTRC, the experimental data of the reactor physics quantities are acquired, such as critical mass, the reactivity worth of simulated control rods and burnable poison rods, the temperature factor of reactivity, power distribution and so on, and the experiment and analysis are advanced. The cores built up in the VHTRC so far were three kinds having different lattice forms and degrees of uranium enrichment. The calculated critical mass was smaller by 1-5 % than the measured values. As to the power distribution and the reactivity worth of burnable poison rods, the prospect of satisfying the required accuracy for the design of the HTTR core was obtained. The experiment using a new core having axially different enrichment degree is planned. (K.I.)

Hydrogen production system coupled with high-temperature gas-cooled reactor (HTTR)

International Nuclear Information System (INIS)

Shiozawa, Shusaku

2003-01-01

On the HTTR program, R and D on nuclear reactor technology and R and D on thermal application technology such as hydrogen production and so on, are advanced. When carrying out power generation and thermal application such as hydrogen production and so on, it is, at first, necessary to supply nuclear heat safely, stably and in low cost, JAERI carries out some R and Ds on nuclear reactor technology using HTTR. In parallel to this, JAERI also carries out R and D for jointing nuclear reactor system with thermal application systems because of no experience in the world on high temperature heat of about 1,000 centigrade supplied by nuclear reactor except power generation, and R and D on thermochemical decomposition method IS process for producing hydrogen from water without exhaust of carbon dioxide. Here were described summaries on R and D on nuclear reactor technology, R and D on jointing technology using HTTR hydrogen production system, R and D on IS process hydrogen production, and comparison hydrogen production with other processes. (G.K.)

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

Effect of ultra high temperature ceramics as fuel cladding materials on the nuclear reactor performance by SERPENT Monte Carlo code

Energy Technology Data Exchange (ETDEWEB)

Korkut, Turgay; Kara, Ayhan; Korkut, Hatun [Sinop Univ. (Turkey). Dept. of Nuclear Energy Engineering

2016-12-15

Ultra High Temperature Ceramics (UHTCs) have low density and high melting point. So they are useful materials in the nuclear industry especially reactor core design. Three UHTCs (silicon carbide, vanadium carbide, and zirconium carbide) were evaluated as the nuclear fuel cladding materials. The SERPENT Monte Carlo code was used to model CANDU, PWR, and VVER type reactor core and to calculate burnup parameters. Some changes were observed at the same burnup and neutronic parameters (keff, neutron flux, absorption rate, and fission rate, depletion of U-238, U-238, Xe-135, Sm-149) with the use of these UHTCs. Results were compared to conventional cladding material zircalloy.

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.)

Transmutation of Thermocouples in Thermal and Fast Nuclear Reactors

International Nuclear Information System (INIS)

Scervini, M.; Rae, C.; Lindley, B.

2013-06-01

Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. Their role is fundamental for the control of current nuclear reactors and for the development of the nuclear technology needed for the implementation of GEN IV nuclear reactors. When used for in-core measurements thermocouples are strongly affected not only by high temperatures, but also by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition in the thermoelements and, as a consequence, a time dependent drift in the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. In this work, undertaken as part of the European project METROFISSION, the change in composition occurring in irradiated thermocouples has been calculated using the software ORIGEN 2.2. Several thermocouples have been considered, including Nickel based thermocouples (type K and type N), Tungsten based thermocouples (W-5%Re vs W-26%Re and W- 3%Re vs W-25%Re), Platinum based thermocouples (type S and Platinum vs Palladium) and Molybdenum vs Niobium thermocouples. The transmutation induced by both thermal flux and fast flux has been calculated. Thermocouples undergo more pronounced transmutation in thermal fluxes rather than in fast fluxes, as the neutron cross section of an element is higher for thermal energies. Nickel based thermocouples have a minimal change in composition, while Platinum based and Tungsten based thermocouples experience a very significant transmutation. The use of coatings deposited on the sheath of a thermocouple has been considered as a mean to reduce the neutron flux the thermoelements inside the thermocouple sheath

Dynamic Complexity Study of Nuclear Reactor and Process Heat Application Integration

Energy Technology Data Exchange (ETDEWEB)

J' Tia Patrice Taylor; David E. Shropshire

2009-09-01

Abstract This paper describes the key obstacles and challenges facing the integration of nuclear reactors with process heat applications as they relate to dynamic issues. The paper also presents capabilities of current modeling and analysis tools available to investigate these issues. A pragmatic approach to an analysis is developed with the ultimate objective of improving the viability of nuclear energy as a heat source for process industries. The extension of nuclear energy to process heat industries would improve energy security and aid in reduction of carbon emissions by reducing demands for foreign derived fossil fuels. The paper begins with an overview of nuclear reactors and process application for potential use in an integrated system. Reactors are evaluated against specific characteristics that determine their compatibility with process applications such as heat outlet temperature. The reactor system categories include light water, heavy water, small to medium, near term high-temperature, and far term high temperature reactors. Low temperature process systems include desalination, district heating, and tar sands and shale oil recovery. High temperature processes that support hydrogen production include steam reforming, steam cracking, hydrogen production by electrolysis, and far-term applications such as the sulfur iodine chemical process and high-temperature electrolysis. A simple static matching between complementary systems is performed; however, to gain a true appreciation for system integration complexity, time dependent dynamic analysis is required. The paper identifies critical issues arising from dynamic complexity associated with integration of systems. Operational issues include scheduling conflicts and resource allocation for heat and electricity. Additionally, economic and safety considerations that could impact the successful integration of these systems are considered. Economic issues include the cost differential arising due to an integrated

Present status of space nuclear reactor

International Nuclear Information System (INIS)

Kaneko, Yoshihiko

1996-01-01

USA and former USSR led space development, and had the experience of launching nuclear reactor satellites. In USA, the research and development of space nuclear reactor were advanced mainly by NASA, and in 1965, the nuclear reactor for power source ''SNAP-10A'' was launched and put on the orbit around the earth. Thereafter, the reactor was started up, and the verifying test at 500 We was successfully carried out. Also for developing the reactor for thermal propulsion, NERVA/ROVER project was done till 1973, and the technological basis was established. The space Exploration Initiative for sending mankind to other solar system planets than the earth is the essential point of the future projects. In former USSR, the ground experiment of the reactor for 800 We power source ''Romashka'', the development of the reactor for 10 kWe power source ''Topaz-1 and 2'', the flight of the artificial satellites, Cosmos 954 and Cosmos 1900, on which nuclear reactors were mounted, and the operation of 33 ocean-monitoring satellites ''RORSAT'' using small fast reactors were carried out. The mission of space development and the nuclear reactors as power source, the engineering of space nuclear reactors, the present status and the trend of space nuclear reactor development, and the investigation by the UN working group on the safety problem of space nuclear reactors are described. (K.I.)

ORTAP: a nuclear steam supply system simulation for the dynamic analysis of high temperature gas cooled reactor transients

International Nuclear Information System (INIS)

Cleveland, J.C.; Hedrick, R.A.; Ball, S.J.; Delene, J.G.

1977-01-01

ORTAP was developed to predict the dynamic behavior of the high temperature gas cooled reactor (HTGR) Nuclear Steam Supply System for normal operational transients and postulated accident conditions. It was developed for the Nuclear Regulatory Commission (NRC) as an independent means of obtaining conservative predictions of the transient response of HTGRs over a wide range of conditions. The approach has been to build sufficient detail into the component models so that the coupling between the primary and secondary systems can be accurately represented and so that transients which cover a wide range of conditions can be simulated. System components which are modeled in ORTAP include the reactor core, a typical reheater and steam generator module, a typical helium circulator and circulator turbine and the turbine generator plant. The major plant control systems are also modeled. Normal operational transients which can be analyzed with ORTAP include reactor start-up and shutdown, normal and rapid load changes. Upset transients which can be analyzed with ORTAP include reactor trip, turbine trip and sudden reduction in feedwater flow. ORTAP has also been used to predict plant response to emergency or faulted conditions such as primary system depressurization, loss of primary coolant flow and uncontrolled removal of control poison from the reactor core

The analysis of thermal-hydraulic performances of nuclear ship reactor

International Nuclear Information System (INIS)

Wakabayashi, Shinshichi; Hamada, Masao

1975-01-01

Thermal-hydraulic performances in the core of nuclear ship reactor was analysed by thermal-hydraulic analyser codes, AMRTC and COBRA-11+DNBCAL. This reactor is of a pressurized water type and incorporates the steam generator within the reactor vessel with the rated power of 330 MWt, which is developed by Nuclear Ship Research Panel Seven (NSR-7) in The Shipbuilding Research Association of Japan. Fuel temperature distributions, coolant temperature distributions, void fractions in coolant and minimum burn out ratio etc. were calculated. Results are as follows; a) The maximum temperature of fuel center is 1,472 0 C that corresponds to 53% as small as the melting point (2,800 0 C). b) Subcooled boiling exists in the core and the maximum void fraction is less than 4%. c) The minimum burn out ratio is not less than the minimum allowable limit of 1.25. It was found from the results of analysis that this reactor was able to be operated wide margin with respect to thermal-hydraulic design limits at the rated power. (auth.)

Irradiation of quench protection diodes at cryogenic temperatures in a nuclear research reactor

International Nuclear Information System (INIS)

Hagedorn, D.; Schoenbacher, H.; Gerstenberg, H.

1996-01-01

Within the framework of the Large Hadron Collider (LHC) R ampersand D programme, CERN and the Department of Physics E21 of the Technical University Munich have established a collaboration to carry out irradiation experiments at liquid helium and liquid nitrogen temperatures on epitaxial diodes for the superconducting magnet protection. Small diode samples of 10 mm wafer diameter from two different manufacturers were submitted to doses of up 50 kGy and neutron fluences up to 1015 n/cm 2 and the degradation of the electrical characteristics was measured versus dose. During irradiation the diodes were submitted to current pulse annealing and after irradiation to thermal annealing. After exposure some diodes show a degradation in forward voltage drop of up to 600 % which, however, can be reduced to about 15 % - 20 % by thermal annealing. The degradation at liquid helium temperature is very similar to the degradation at liquid nitrogen temperature. These degradations of electrical characteristics during the short term irradiation in a nuclear reactor are compared with degradations during long term irradiation in an accelerator environment at liquid nitrogen temperature

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)

Advanced gas cooled nuclear reactor materials evaluation and development program

International Nuclear Information System (INIS)

1977-01-01

Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed

Data-processing program from the operating modes of the nuclear reactor (P0DER)

International Nuclear Information System (INIS)

Totev, T.L.; Boyadzhiev, A.I.

1981-01-01

A program PODER for processing data from the operating modes of the reactors taking into account the effects of corrosion, hydration, and deformation of the nuclear reactor fuel element sheathing, the formation of the corrosion product deposits, the change in the geometric dimensions of the nuclear reactor fuel element due to the temperature deformation, as well as the various gas fillers, are elaborated. The ''hot channel'' method determining the reliability of the system is realized. The basic equations describing the thermohydraulic processes in nuclear reactors are solved by the finite difference method. Approximations are carried out with the approach of least squares. The temperature distribution versus the zirconium sheathing height is computed for the case of WWER-440 type reactors. The advantages of the proposed program P0DER are discussed

Conceptual design of nuclear fusion power reactor DREAM. Reactor structures and remote maintenance

International Nuclear Information System (INIS)

Nishio, Satoshi; Seki, Yasushi; Ueda, Shuzo; Kurihara, Ryoichi; Adachi, Junichi; Yamazaki, Seiichiro; Hashimoto, Toshiyuki.

1997-01-01

Nuclear fusion reactors are required to be able to compete another energy sources in economy, reliability, safety and environmental integrity for commercial use. In the DREAM (DRastically EAsy Maintenance) reactor, a very low activated material of SiC/SiC composite has been introduced for the structural material, a reactor configuration for very easy maintenance and the helium gas of a high temperature for the cooling system, and hence DREAM has been proven to be very attractively as the commercial power reactor due to the high availability and efficiency of the plant and minimization of radioactive wastes. (author)

Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

Energy Technology Data Exchange (ETDEWEB)

Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

2009-10-01

High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next

Fast reactors in nuclear power

Energy Technology Data Exchange (ETDEWEB)

Kazachkovskii, O

1981-02-01

The possible applications are discussed of fast reactor nuclear power plants. Basic differences are explained in fast and thermal reactors, mainly with a view to nuclear fuel utilization. Discussed in more detail are the problems of nuclear fuel reproduction and the nost important technical problems of fast reactors. Flow charts are shown of heat transfer for fast reactors BN-350 (loop design) and BN-600 (integral coolant circuit design). Main specifications are given for demonstration and power fast reactors in operation, under construction and in project-stage.

Nuclear reactors; graphical symbols

International Nuclear Information System (INIS)

1987-11-01

This standard contains graphical symbols that reveal the type of nuclear reactor and is used to design graphical and technical presentations. Distinguishing features for nuclear reactors are laid down in graphical symbols. (orig.) [de

Assessment of very high-temperature reactors in process application. Appendix I. Evaluation of the reactor system

International Nuclear Information System (INIS)

Jones, J.E. Jr.; Spiewak, I.

1976-12-01

In April 1974, the U.S. Atomic Energy Commission [now the Energy Research and Development Administration (ERDA)] authorized General Atomic Company, General Electric Company, and Westinghouse Electric Corp., Astronuclear Laboratory, to assess the available technology for producing heat using very high-temperature nuclear reactors. An evaulation of these studies and of the technical and economic potential of very high-temperature reactors (VHTR) is presented. The VHTR is a helium-cooled graphite-moderated reactor. The concepts and technology are evaluated for producing process stream temperatures of 649, 760, 871, 982, and 1093 0 C (1200, 1400, 1600, 1800, and 2000 0 F). There are a number of large industrial process heat applications that could utilize the VHTR

Nuclear fuels for very high temperature applications

International Nuclear Information System (INIS)

Lundberg, L.B.; Hobbins, R.R.

1992-01-01

The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO 2 or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures

Guidebook to nuclear reactors

International Nuclear Information System (INIS)

Nero, A.V. Jr.

1976-05-01

A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen

Hydrogen production by water-splitting using heat supplied by a high-temperature reactor

International Nuclear Information System (INIS)

Courvoisier, P.; Rastoin, J.; Titiliette, Z.

1976-01-01

Some aspects of the use of heat of nuclear origin for the production of hydrogen by water-splitting are considered. General notions pertaining to the yield of chemical cycles are discussed and the heat balance corresponding to two specific processes is evaluated. The possibilities of high temperature reactors, with respect to the coolant temperature levels, are examined from the standpoint of core design and technology of some components. Furthermore, subject to a judicious selection of their characteristics, these reactors can lead to excellent use of nuclear fuel. The coupling of the nuclear reactor with the chemical plant by means of a secondary helium circuit gives rise to the design of an intermediate heat exchanger, which is an important component of the overall installation. (orig.) [de

High temperature fast reactor for hydrogen production in Brazil

International Nuclear Information System (INIS)

Nascimento, Jamil A. do; Ono, Shizuca; Guimaraes, Lamartine N.F.

2008-01-01

The main nuclear reactors technology for the Generation IV, on development phase for utilization after 2030, is the fast reactor type with high temperature output to improve the efficiency of the thermo-electric conversion process and to enable applications of the generated heat in industrial process. Currently, water electrolysis and thermo chemical cycles using very high temperature are studied for large scale and long-term hydrogen production, in the future. With the possible oil scarcity and price rise, and the global warming, this application can play an important role in the changes of the world energy matrix. In this context, it is proposed a fast reactor with very high output temperature, ∼ 1000 deg C. This reactor will have a closed fuel cycle; it will be cooled by lead and loaded with nitride fuel. This reactor may be used for hydrogen, heat and electricity production in Brazil. It is discussed a development strategy of the necessary technologies and some important problems are commented. The proposed concept presents characteristics that meet the requirements of the Generation IV reactor class. (author)

Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

2008-08-01

Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of

High temperature heat exchange: nuclear process heat applications

International Nuclear Information System (INIS)

Vrable, D.L.

1980-09-01

The unique element of the HTGR system is the high-temperature operation and the need for heat exchanger equipment to transfer nuclear heat from the reactor to the process application. This paper discusses the potential applications of the HTGR in both synthetic fuel production and nuclear steel making and presents the design considerations for the high-temperature heat exchanger equipment

An overview of future sustainable nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Poullikkas, Andreas [Electricity Authority of Cyprus, P.O. Box 24506, 1399 Nicosia (Cyprus)

2013-07-01

In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA). In contrast, generation III reactors, which are an evolution of generation II reactors, incorporate passive or inherent safety features that require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. Today, partly due to the high capital cost of large power reactors generating electricity and partly due to the consideration of public perception, there is a shift towards the development of smaller units. These may be built independently or as modules in a larger complex, with capacity added incrementally as required. Small reactors most importantly benefit from reduced capital costs, simpler units and the ability to produce power away from main grid systems. These factors combined with the ability of a nuclear power plant to use process heat for co-generation, make the small reactors an attractive option. Generally, modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production and reduced installation costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction. Generation III+ designs are generally extensions of the generation III concept, which include advanced passive safety features. These designs can maintain the safe state without the use of any active control components. Generation IV reactors, which are future designs that are currently under research and development, will tend to have closed

Molten salt reactors and possible scenarios for future nuclear power deployment

International Nuclear Information System (INIS)

Merle-Lucotte, E.; Mathieu, L.; Heuer, D.; Loiseaux, J. M.; Billebaud, A.; Brissot, R.; David, S.; Garzenne, C.; Laulan, O.; Le Brun, C.; Lecarpentier, D.; Liatard, E.; Meplan, O.; Michel-Sendis, F.; Nuttin, A.; Perdu, F.

2004-01-01

An important fraction of the nature energy demand may be satisfied by nuclear power. In this context, the possibilities of worldwide nuclear deployment are studied. We are convinced that the Molten Salt Reactors may play a central role in this deployment. The Molten Salt Reactor needs to be coupled to a reprocessing unit in order to extract the Fission Products which poison the core. The efficiency of this reprocessing has a crucial influence on reactor behavior especially for the breeding ratio. The Molten Salt Breeder Reactor project was based on an intensive reprocessing for high breeding purposes. A new concept of Thorium Molten Salt Reactor is presented here. Including this new concept in the worldwide nuclear deployment, to satisfy these power needs, we consider three typical scenarios, based on three reactor types: Pressurized Water Reactor, Fast Neutron Reactor and Thorium Molten Salt Reactor. The aim of this paper is to demonstrate, in a first hand that a Thorium Molten Salt Reactor can be realistic, with correct temperature coefficients and at least iso-breeder with slow reprocessing and new geometry; on the other hand that such Molten Salt Reactors enable a successful nuclear deployment, while minimizing fuel and waste management problems. (authors)

Nuclear reactor safety research in Idaho

International Nuclear Information System (INIS)

Zeile, H.J.

1983-01-01

Detailed information about the performance of nuclear reactor systems, and especially about the nuclear fuel, is vital in determining the consequences of a reactor accident. Fission products released from the fuel during accidents are the ultimate safety concern to the general public living in the vicinity of a nuclear reactor plant. Safety research conducted at the Idaho National Engineering Laboratory (INEL) in support of the U.S. Nuclear Regulatory Commission (NRC) has provided the NRC with detailed data relating to most of the postulated nuclear reactor accidents. Engineers and scientists at the INEL are now in the process of gathering data related to the most severe nuclear reactor accident - the core melt accident. This paper describes the focus of the nuclear reactor safety research at the INEL. The key results expected from the severe core damage safety research program are discussed

An introduction to the engineering of fast nuclear reactors

CERN Document Server

Judd, Anthony M

2014-01-01

An invaluable resource for both graduate-level engineering students and practising nuclear engineers who want to expand their knowledge of fast nuclear reactors, the reactors of the future! This book is a concise yet comprehensive introduction to all aspects of fast reactor engineering. It covers topics including neutron physics; neutron flux spectra; flux distribution; Doppler and coolant temperature coefficients; the performance of ceramic and metal fuels under irradiation, structural changes, and fission-product migration; the effects of irradiation and corrosion on structural materials, irradiation swelling; heat transfer in the reactor core and its effect on core design; coolants including sodium and lead-bismuth alloy; coolant circuits; pumps; heat exchangers and steam generators; and plant control. The book includes new discussions on lead-alloy and gas coolants, metal fuel, the use of reactors to consume radioactive waste, and accelerator-driven subcritical systems.

Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

International Nuclear Information System (INIS)

Schikorr, W.M.

In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

Advanced nuclear reactor safety issues and research needs

International Nuclear Information System (INIS)

2002-01-01

On 18-20 February 2002, the OECD Nuclear Energy Agency (NEA) organised, with the co-sponsorship of the International Atomic Energy Agency (IAEA) and in collaboration with the European Commission (EC), a Workshop on Advanced Nuclear Reactor Safety Issues and Research Needs. Currently, advanced nuclear reactor projects range from the development of evolutionary and advanced light water reactor (LWR) designs to initial work to develop even further advanced designs which go beyond LWR technology (e.g. high-temperature gas-cooled reactors and liquid metal-cooled reactors). These advanced designs include a greater use of advanced technology and safety features than those employed in currently operating plants or approved designs. The objectives of the workshop were to: - facilitate early identification and resolution of safety issues by developing a consensus among participating countries on the identification of safety issues, the scope of research needed to address these issues and a potential approach to their resolution; - promote the preservation of knowledge and expertise on advanced reactor technology; - provide input to the Generation IV International Forum Technology Road-map. In addition, the workshop tried to link advancement of knowledge and understanding of advanced designs to the regulatory process, with emphasis on building public confidence. It also helped to document current views on advanced reactor safety and technology, thereby contributing to preserving knowledge and expertise before it is lost. (author)

Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

International Nuclear Information System (INIS)

Glasstone, S.; Sesonske, A.

1994-01-01

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials

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.)

Modeling the high-temperature gas-cooled reactor process heat plant: a nuclear to chemical conversion process

International Nuclear Information System (INIS)

Pfremmer, R.D.; Openshaw, F.L.

1982-05-01

The high-temperature heat available from the High-Temperature Gas-Cooled Reactor (HTGR) makes it suitable for many process applications. One of these applications is a large-scale energy production plant where nuclear energy is converted into chemical energy and stored for industrial or utility applications. This concept combines presently available nuclear HTGR technology and energy conversion chemical technology. The design of this complex plant involves questions of interacting plant dynamics and overall plant control. This paper discusses how these questions were answered with the aid of a hybrid computer model that was developed within the time-frame of the conceptual design studies. A brief discussion is given of the generally good operability shown for the plant and of the specific potential problems and their anticipated solution. The paper stresses the advantages of providing this information in the earliest conceptual phases of the design

The analysis for inventory of experimental reactor high temperature gas reactor type

International Nuclear Information System (INIS)

Sri Kuntjoro; Pande Made Udiyani

2016-01-01

Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

International Nuclear Information System (INIS)

Glasstone, S.; Sesonske, A.

1994-01-01

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future

Nuclear reactor

International Nuclear Information System (INIS)

Batheja, P.; Huber, R.; Rau, P.

1985-01-01

Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

Effective neutron temperature measurements in well moderated reactor by the reactivity coefficient method

International Nuclear Information System (INIS)

Raisic, N.; Klinc, T.

1968-11-01

The ratio of the reactivity changes of a nuclear reactor produced by successive introduction of two different neutron absorbers in the reactor core, has been measured and information on effective neutron temperature at a particular point obtained. Boron was used as a l/v absorber and cadmium as an absorber sensiti ve to neutron temperature. Effective neutron temperature distribution has been deduced by moving absorbers across the reactor core and observing the corresponding reactivity changes. (author)

A coupled nuclear reactor thermal energy storage system for enhanced load following operation

International Nuclear Information System (INIS)

Alameri, Saeed A.; King, Jeffrey C.

2013-01-01

Nuclear power plants operate most economically at a constant power level, providing base load electric power. In an energy grid containing a high fraction of renewable power sources, nuclear reactors may be subject to significantly variable power demands. These variable power demands can negatively impact the effective capacity factor of the reactor and result in severe economic penalties. Coupling a nuclear reactor to a large thermal energy storage block will allow the reactor to better respond to variable power demands. In the system described in this paper, a Prismatic core Advanced High Temperature Reactor supplies constant power to a lithium chloride molten salt thermal energy storage block that provides thermal power as needed to a closed Brayton cycle energy conversion system. During normal operation, the thermal energy storage block stores thermal energy during the night for use in the times of peak demand during the day. In this case, the nuclear reactor stays at a constant thermal power level. After a loss of forced circulation, the reactor reaches a shut down state in less than half an hour and the average fuel, graphite and coolant temperatures remain well within the design limits over the duration of the transient, demonstrating the inherent safety of the coupled system. (author)

Role of temperature in the radiation stability of yttria stabilized zirconia under swift heavy ion irradiation: A study from the perspective of nuclear reactor applications

Science.gov (United States)

Kalita, Parswajit; Ghosh, Santanu; Sattonnay, Gaël; Singh, Udai B.; Grover, Vinita; Shukla, Rakesh; Amirthapandian, S.; Meena, Ramcharan; Tyagi, A. K.; Avasthi, Devesh K.

2017-07-01

The search for materials that can withstand the harsh radiation environments of the nuclear industry has become an urgent challenge in the face of ever-increasing demands for nuclear energy. To this end, polycrystalline yttria stabilized zirconia (YSZ) pellets were irradiated with 80 MeV Ag6+ ions to investigate their radiation tolerance against fission fragments. To better simulate a nuclear reactor environment, the irradiations were carried out at the typical nuclear reactor temperature (850 °C). For comparison, irradiations were also performed at room temperature. Grazing incidence X-ray diffraction and Raman spectroscopy measurements reveal degradation in crystallinity for the room temperature irradiated samples. No bulk structural amorphization was however observed, whereas defect clusters were formed as indicated by transmission electron microscopy and supported by thermal spike simulation results. A significant reduction of the irradiation induced defects/damage, i.e., improvement in the radiation tolerance, was seen under irradiation at 850 °C. This is attributed to the fact that the rapid thermal quenching of the localized hot molten zones (arising from spike in the lattice temperature upon irradiation) is confined to 850 °C (i.e., attributed to the resistance inflicted on the rapid thermal quenching of the localized hot molten zones by the high temperature of the environment) thereby resulting in the reduction of the defects/damage produced. Our results present strong evidence for the applicability of YSZ as an inert matrix fuel in nuclear reactors, where competitive effects of radiation damage and dynamic thermal healing mechanisms may lead to a strong reduction in the damage production and thus sustain its physical integrity.

Nuclear reactor containment device

International Nuclear Information System (INIS)

Ichiki, Tadaharu.

1980-01-01

Purpose: To reduce the volume of a containment shell and decrease the size of a containment equipment for BWR type reactors by connecting the containment shell and a suppression pool with slanted vent tubes to thereby shorten the vent tubes. Constitution: A pressure vessel containing a reactor core is installed at the center of a building and a containment vessel for the nuclear reactor that contains the pressure vessel forms a cabin. To a building situated below the containment shell, is provided a suppression chamber in which cooling water is charged to form a suppression pool. The suppression pool is communicated with vent tubes that pass through the partition wall of the containment vessel. The vent tubes are slanted and their lower openings are immersed in coolants. Therefore, if accident is resulted and fluid at high temperature and high pressure is jetted from the pressure vessel, the jetting fluid is injected and condensated in the cooling water. (Moriyama, K.)

Microstructural control and high temperature mechanical property of ferritic/martensitic steels for nuclear reactor application

International Nuclear Information System (INIS)

Adetunji, G.J.

1991-04-01

The materials under study are 9-12% Cr ferritic/martensitic steels, alternative candidate materials for application in core components of nuclear power reactors. This work involves (1) Investigation of high temperature fracture mechanism during slow tensile and limited creep testing at 600 o C (2) Extensive study of solute element segregation both theoretically and experimentally (3) Investigation of effects by thermal ageing and irradiation on microstructural developments in relation to high temperature mechanical behaviour. From (1) the results obtained indicate that the important microstructural characteristics controlling the fracture of 9-12% Cr ferritic/martensitic steels at high temperature are (a) solute segregation to inclusion-matrix interfaces (b) hardness of the martensitic matrix and (c) carbide particle size distribution. From (2) the results indicate a strong concentration gradient of silicon and molybdenum near lath packet boundaries for certain quenching rates from the austenitizing temperature. From (3) high temperature tensile data were obtained for irradiated samples with thermally aged ones as control. (author)

Comparative Analysis of Hydrogen Production Methods with Nuclear Reactors

International Nuclear Information System (INIS)

Morozov, Andrey

2008-01-01

Hydrogen is highly effective and ecologically clean fuel. It can be produced by a variety of methods. Presently the most common are through electrolysis of water and through the steam reforming of natural gas. It is evident that the leading method for the future production of hydrogen is nuclear energy. Several types of reactors are being considered for hydrogen production, and several methods exist to produce hydrogen, including thermochemical cycles and high-temperature electrolysis. In the article the comparative analysis of various hydrogen production methods is submitted. It is considered the possibility of hydrogen production with the nuclear reactors and is proposed implementation of research program in this field at the IPPE sodium-potassium eutectic cooling high temperature experimental facility (VTS rig). (authors)

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

Safety Philosophy in Process Heat Plants Coupled to High Temperature Reactors

International Nuclear Information System (INIS)

Brown, Nicholas R.; Revankar, Shripad T.

2011-01-01

With the future availability of fossil fuel resources in doubt, high temperature nuclear reactors have the potential to be an important technology in the near term. Due to a high coolant outlet temperature, high temperature reactors (HTR) can be used to drive chemical plants that directly utilize process heat. Additionally, the high temperature improves the thermodynamic efficiency of the energy utilization. Many applications of high temperature reactors exist as a thermal driving vector for endothermic chemical process plants. Hydrogen generation using the General Atomics (GA) sulfur iodine (SI) cycle is one promising application of high temperature nuclear heat. The main chemical reactions in the SI cycle are: 1. I 2 +SO 2 + 2H 2 O → 2HI + H 2 SO 4 (Bunsen reaction) 2. H 2 SO 4 → H 2 O + SO 2 + 1/2O 2 (Sulfuric acid decomposition) 3. 2HI → H 2 + I 2 (Hydrogen Iodide decomposition). With the exception of hydrogen and oxygen, all relevant reactants are recycled within the process. However, there are many unresolved safety and operational issues related to implementation of such a coupled plant

Reactor physics analysis of the pin-cell Doppler effect in a thermal nuclear reactor

International Nuclear Information System (INIS)

Kruijf, W.J.M. de.

1995-01-01

This report has also been published as a PhD thesis. It deals with the Doppler effect in thermal nuclear reactors. Especially the behaviour of the reactor in transient conditions is an important issue. During such a transient the radial temperature profile in a fuel pin changes. In this PhD research effective fuel temperatures have been calculated for arbitrary temperature profiles in the fuel pin with the improved slowing-down code ROLAIDS-CPM. A general expression for the effective fuel temperature in a specific fuel pin is found by defining this effective fuel temperature as a weighted sum of the temperatures in different radial fuel zones. Also, the radial power profile in a fuel pin has been calculated by performing detailed burnup calculations, which agree very well with experimental data. (orig.)

Application of heat pipes in nuclear reactors for passive heat removal

Energy Technology Data Exchange (ETDEWEB)

Haque, Z.; Yetisir, M., E-mail: haquez@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2013-07-01

This paper introduces a number of potential heat pipe applications in passive (i.e., not requiring external power) nuclear reactor heat removal. Heat pipes are particularly suitable for small reactors as the demand for heat removal is significantly less than commercial nuclear power plants, and passive and reliable heat removal is required. The use of heat pipes has been proposed in many small reactor designs for passive heat removal from the reactor core. This paper presents the application of heat pipes in AECL's Nuclear Battery design, a small reactor concept developed by AECL. Other potential applications of heat pipes include transferring excess heat from containment to the atmosphere by integrating low-temperature heat pipes into the containment building (to ensure long-term cooling following a station blackout), and passively cooling spent fuel bays. (author)

Coal gasification coal by steam using process heat from high-temperature nuclear reactors

International Nuclear Information System (INIS)

Heek, K.H. van; Juentgen, H.; Peters, W.

1982-01-01

This paper outlines the coal gasification process using a high-temperature nuclear reactor as a source of the process heat needed. Compared to conventional gasification processes coal is saved by 30-40%, coal-specific emissions are reduced and better economics of gas production are achieved. The introductory chapter deals with motives, aims and tasks of the development, followed by an explanation of the status of investigations, whereby especially the results of a semi-technical pilot plant operated by Bergbau-Forschung are given. Furthermore, construction details of a full-scale commercial gasifier are discussed, including the development of suitable alloys for the heat exchanger. Moreover problems of safety, licensing and economics of future plants have been investigated. (orig.) [de

Application of finite element numerical technique to nuclear reactor geometries

Energy Technology Data Exchange (ETDEWEB)

Rouai, N M [Nuclear engineering department faculty of engineering Al-fateh universty, Tripoli (Libyan Arab Jamahiriya)

1995-10-01

Determination of the temperature distribution in nuclear elements is of utmost importance to ensure that the temperature stays within safe limits during reactor operation. This paper discusses the use of Finite element numerical technique (FE) for the solution of the two dimensional heat conduction equation in geometries related to nuclear reactor cores. The FE solution stats with variational calculus which considers transforming the heat conduction equation into an integral equation I(O) and seeks a function that minimizes this integral and hence gives the solution to the heat conduction equation. In this paper FE theory as applied to heat conduction is briefly outlined and a 2-D program is used to apply the theory to simple shapes and to two gas cooled reactor fuel elements. Good results are obtained for both cases with reasonable number of elements. 7 figs.

Application of finite element numerical technique to nuclear reactor geometries

International Nuclear Information System (INIS)

Rouai, N. M.

1995-01-01

Determination of the temperature distribution in nuclear elements is of utmost importance to ensure that the temperature stays within safe limits during reactor operation. This paper discusses the use of Finite element numerical technique (FE) for the solution of the two dimensional heat conduction equation in geometries related to nuclear reactor cores. The FE solution stats with variational calculus which considers transforming the heat conduction equation into an integral equation I(O) and seeks a function that minimizes this integral and hence gives the solution to the heat conduction equation. In this paper FE theory as applied to heat conduction is briefly outlined and a 2-D program is used to apply the theory to simple shapes and to two gas cooled reactor fuel elements. Good results are obtained for both cases with reasonable number of elements. 7 figs

Refuelling nuclear reactors

International Nuclear Information System (INIS)

Stacey, J.; Webb, J.; White, W.P.; McLaren, N.H.

1981-01-01

An improved nuclear reactor refuelling machine is described which can be left in the reactor vault to reduce the off-load refuelling time for the reactor. The system comprises a gripper device rangeable within a tubular chute, the gripper device being movable by a pantograph. (U.K.)

Nuclear reactor instrumentation at research reactor renewal

International Nuclear Information System (INIS)

Baers, B.; Pellionisz, P.

1981-10-01

The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

Feasibility study of high temperature reactor utilization in Czech Republic after 2025

Energy Technology Data Exchange (ETDEWEB)

Losa, Evžen, E-mail: evzen.losa@fjfi.cvut.cz [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Nuclear Reactors (Czech Republic); Heřmanský, Bedřich; Kobylka, Dušan; Rataj, Jan; Sklenka, Ľubomír [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Nuclear Reactors (Czech Republic); Souček, Václav; Kohout, Petr [AZIN CZ, s.r.o., Hanusova 3, 140 00 Praha 4 (Czech Republic)

2014-05-01

High temperature reactors (HTRs) were examined as an option to intended future broadening of the nuclear energy production in Czech Republic. The known qualities as the inherent safety, high thermal utilization and non-electrical applications have been assessed in years 2009–2011 during the survey funded by Czech Ministry of Industry and Trade. The survey of high temperature reactors with spherical fuel was initiated by reason of mature state of the art of this technology type in South Africa and in China, where in both countries pilot plants were planned. Unfortunately, the global financial crisis caused the decision of stopping the governmental support in South African programme was made. In China, however, the development still continues. Czech Republic has almost 60 years nuclear research history and the knowledge of operation of gas cooled and heavy water moderated reactor has been gained in the past. Nevertheless, the design of light water reactors was more developed in former Soviet Union, which provided Czech scientists by initial knowledge base; hence the research has been reoriented to this technology. But, the demands on future nuclear reactors application are still growing and the same or even higher living standard of next generations have to be taken into consideration. Therefore the systems, which can produce more energy and less waste, are getting into foreground of interest of Czech decision makers. The high temperature reactor technology seems to be the successful representative of the GEN IV reactor types, which will be operated commercially in the near future. The broad spectrum of utilization enables this system to be an option after 2030, when the electricity demand is planned to be covered from about 50% by nuclear in our country.

Feasibility study of high temperature reactor utilization in Czech Republic after 2025

International Nuclear Information System (INIS)

Losa, Evžen; Heřmanský, Bedřich; Kobylka, Dušan; Rataj, Jan; Sklenka, Ľubomír; Souček, Václav; Kohout, Petr

2014-01-01

High temperature reactors (HTRs) were examined as an option to intended future broadening of the nuclear energy production in Czech Republic. The known qualities as the inherent safety, high thermal utilization and non-electrical applications have been assessed in years 2009–2011 during the survey funded by Czech Ministry of Industry and Trade. The survey of high temperature reactors with spherical fuel was initiated by reason of mature state of the art of this technology type in South Africa and in China, where in both countries pilot plants were planned. Unfortunately, the global financial crisis caused the decision of stopping the governmental support in South African programme was made. In China, however, the development still continues. Czech Republic has almost 60 years nuclear research history and the knowledge of operation of gas cooled and heavy water moderated reactor has been gained in the past. Nevertheless, the design of light water reactors was more developed in former Soviet Union, which provided Czech scientists by initial knowledge base; hence the research has been reoriented to this technology. But, the demands on future nuclear reactors application are still growing and the same or even higher living standard of next generations have to be taken into consideration. Therefore the systems, which can produce more energy and less waste, are getting into foreground of interest of Czech decision makers. The high temperature reactor technology seems to be the successful representative of the GEN IV reactor types, which will be operated commercially in the near future. The broad spectrum of utilization enables this system to be an option after 2030, when the electricity demand is planned to be covered from about 50% by nuclear in our country

Spatial effects on the fluctuations of a nuclear power reactor

International Nuclear Information System (INIS)

Salinas-Rodriguez, E.; Rodriguez, R.F.; Wio, H.S.

1990-01-01

The effects of spatial inhomogeneities in a nuclear system are studied by using the compounding moments method. In particular, the neutron density and temperature equilibrium correlation functions are explicitly calculated for a realistic linearized nuclear reactor model described in terms of a master equation. (author)

In-reactor testing of the closed cycle gas core reactor---the nuclear light bulb concept

International Nuclear Information System (INIS)

Gauntt, R.O.; Slutz, S.A.; Harms, G.A.; Latham, T.S.; Roman, W.C.; Rodgers, R.J.

1993-01-01

The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (>1800 s) and thrust (>445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (∼4000 K). The following paper describes analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented here include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRR. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRR for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept

Ceramics as nuclear reactor fuels

International Nuclear Information System (INIS)

Reeve, K.D.

1975-01-01

Ceramics are widely accepted as nuclear reactor fuel materials, for both metal clad ceramic and all-ceramic fuel designs. Metal clad UO 2 is used commercially in large tonnages in five different power reactor designs. UO 2 pellets are made by familiar ceramic techniques but in a reactor they undergo complex thermal and chemical changes which must be thoroughly understood. Metal clad uranium-plutonium dioxide is used in present day fast breeder reactors, but may eventually be replaced by uranium-plutonium carbide or nitride. All-ceramic fuels, which are necessary for reactors operating above about 750 0 C, must incorporate one or more fission product retentive ceramic coatings. BeO-coated BeO matrix dispersion fuels and silicate glaze coated UO 2 -SiO 2 have been studied for specialised applications, but the only commercial high temperature fuel is based on graphite in which small fuel particles, each coated with vapour deposited carbon and silicon carbide, are dispersed. Ceramists have much to contribute to many aspects of fuel science and technology. (author)

Nuclear data for fission reactor core design and safety analysis: Requirements and status of accuracy of nuclear data

International Nuclear Information System (INIS)

Rowlands, J.L.

1984-01-01

The types of nuclear data required for fission reactor design and safety analysis, and the ways in which the data are represented and approximated for use in reactor calculations, are summarised first. The relative importance of different items of nuclear data in the prediction of reactor parameters is described and ways of investigating the accuracy of these data by evaluating related integral measurements are discussed. The use of sensitivity analysis, together with estimates of the uncertainties in nuclear data and relevant integral measurements, in assessing the accuracy of prediction of reactor parameters is described. The inverse procedure for deciding nuclear data requirements from the target accuracies for prediction of reactor parameters follows on from this. The need for assessments of the uncertainties in nuclear data evaluations and the form of the uncertainty information is discussed. The status of the accuracies of predictions and nuclear data requirements are then summarised. The reactor parameters considered include: (a) Criticality conditions, conversion and burn-up effects. (b) Energy production and deposition, decay heating, irradiation damage, dosimetry and induced radioactivity. (c) Kinetics characteristics and control, including temperature, power and coolant density coefficients, delayed neutrons and control absorbers. (author)

Secondary heat exchanger design and comparison for advanced high temperature reactor

International Nuclear Information System (INIS)

Sabharwall, P.; Kim, E. S.; Siahpush, A.; McKellar, M.; Patterson, M.

2012-01-01

Next generation nuclear reactors such as the advanced high temperature reactor (AHTR) are designed to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. This study considers two different types of heat exchangers - helical coiled heat exchanger and printed circuit heat exchanger - as possible options for the AHTR secondary heat exchangers with distributed load analysis and comparison. Comparison is provided for all different cases along with challenges and recommendations. (authors)

Systems and methods for enhancing isolation of high-temperature reactor containments

Energy Technology Data Exchange (ETDEWEB)

Peterson, Per F.

2017-09-26

A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.

Nuclear reactor internals arrangement

International Nuclear Information System (INIS)

Frisch, E.; Andrews, H.N.

1976-01-01

A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

Indian advanced nuclear reactors

International Nuclear Information System (INIS)

Saha, D.; Sinha, R.K.

2005-01-01

For sustainable development of nuclear energy, a number of important issues like safety, waste management, economics etc. are to be addressed. To do this, a number of advanced reactor designs as well as fuel cycle technologies are being pursued worldwide. The advanced reactors being developed in India are the AHWR and the CHTR. Both the reactors use thorium based fuel and have many passive features. This paper describes the Indian advanced reactors and gives a brief account of the international initiatives for the sustainable development of nuclear energy. (author)

Licensing of nuclear reactor operators

International Nuclear Information System (INIS)

1979-09-01

Recommendations are presented for the licensing of nuclear reactor operators in units licensed according to the legislation in effect. They apply to all physical persons designated by the Operating Organization of the nuclear reactor or reactors to execute any of the following functional activities: a) to manipulate the controls of a definite reactor b) to direct the authorized activities of the reactor operators licesed according to the present recommendations. (F.E.) [pt

Sodium-cooled nuclear reactor

International Nuclear Information System (INIS)

Hammers, H.W.

1982-01-01

The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

Advanced Nuclear Reactor Concepts for China

International Nuclear Information System (INIS)

Knoche, D.; Sassen, F.; Tietsch, W.; Yujie, Dong; Li, Cao

2008-01-01

China is one of the fastest growing economies in the world. With 1.3 billion people China also has the largest population worldwide. The growing economy, the migration of people from rural areas to cities and the augmentation in living standard will drive the energy demand of China in the coming decades. At present the installed electrical power is about 500 GW. In the years 2004 and 2005 the added electrical capacity was around 60 GW per year. Chinas primary energy demand is covered mainly by the use of coal. Coal also will remain the main energy source in the coming decades in China. Nevertheless taking into account more and more environmental aspects and the goal to reduce dependencies on energy imports a better energy mix strategy is planed to change including at an increasing level the renewable and nuclear option. Present the nuclear park is characterised by a large variety of different types of reactors. With the AP-1000, EPR and the gas-cooled High Temperature Reactor (HTR) the spectrum of different reactor types will be further enlarged. (authors)

Advanced Nuclear Reactor Concepts for China

Energy Technology Data Exchange (ETDEWEB)

Knoche, D.; Sassen, F.; Tietsch, W. [Westinghouse Electric Germany, Postfach 10 05 63, 68140 Mannheim (Germany); Yujie, Dong; Li, Cao [INET, Tsinghua University, 100084 Beijing (China)

2008-07-01

China is one of the fastest growing economies in the world. With 1.3 billion people China also has the largest population worldwide. The growing economy, the migration of people from rural areas to cities and the augmentation in living standard will drive the energy demand of China in the coming decades. At present the installed electrical power is about 500 GW. In the years 2004 and 2005 the added electrical capacity was around 60 GW per year. Chinas primary energy demand is covered mainly by the use of coal. Coal also will remain the main energy source in the coming decades in China. Nevertheless taking into account more and more environmental aspects and the goal to reduce dependencies on energy imports a better energy mix strategy is planed to change including at an increasing level the renewable and nuclear option. Present the nuclear park is characterised by a large variety of different types of reactors. With the AP-1000, EPR and the gas-cooled High Temperature Reactor (HTR) the spectrum of different reactor types will be further enlarged. (authors)

Recirculation system for nuclear reactors

International Nuclear Information System (INIS)

Braun, H. E.; Dollard, W. J.; Tower, S. N.

1980-01-01

A recirculation system for use in pressurized water nuclear reactors to increase the output temperature of the reactor coolant, thereby achieving a significant improvement in plant efficiency without exceeding current core design limits. A portion of the hot outlet coolant is recirculated to the inlets of the peripheral fuel assemblies which operate at relatively low power levels. The outlet temperature from these peripheral fuel assemblies is increased to a temperature above that of the average core outlet. The recirculation system uses external pumps and introduces the hot recirculation coolant to the free space between the core barrel and the core baffle, where it flows downward and inward to the inlets of the peripheral fuel assemblies. In the unlikely event of a loss of coolant accident, the recirculation system flow path through the free space and to the inlets of the fuel assemblies is utilized for the injection of emergency coolant to the lower vessel and core. During emergency coolant injection, the emergency coolant is prevented from bypassing the core through the recirculation system by check valves inserted into the recirculation system piping

Nuclear reactor simulator

International Nuclear Information System (INIS)

Baptista, Vinicius Damas

1996-01-01

The Nuclear Reactor Simulator was projected to help the basic training in the formation of the Nuclear Power Plants operators. It gives the trainee the opportunity to see the nuclear reactor dynamics. It's specially indicated to be used as the support tool to NPPT (Nuclear Power Preparatory Training) from NUS Corporation. The software was developed to Intel platform (80 x 86, Pentium and compatible ones) working under the Windows operational system from Microsoft. The program language used in development was Object Pascal and the compiler used was Delphi from Borland. During the development, computer algorithms were used, based in numeric methods, to the resolution of the differential equations involved in the process. (author)

Baffle-former arrangement for nuclear reactor vessel internals

International Nuclear Information System (INIS)

Machado, O.J.; Berringer, R.T.

1978-01-01

A baffle-former arrangement for the reactor vessel internals of a nuclear reactor is described. The arrangement includes positioning of formers at the same elevations as the fuel assembly grids, and positioning flow holes in the baffle plates directly beneath selected former grid elevations. The arrangement reduces detrimental cross flows, maintains proper core barrel and baffle temperatures, and alleviates the potential of overpressurization within the baffle-former assembly under assumed major accident conditions

Solid-Core Heat-Pipe Nuclear Batterly Type Reactor

International Nuclear Information System (INIS)

Ehud Greenspan

2008-01-01

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE 400 space nuclear reactor core, the HPENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The core is oriented horizontally and has a square rather cylindrical cross section for effective heat transfer. The HPs extend from the two axial reflectors in which the fission gas plena are embedded and transfer heat to an intermediate coolant that flows by natural-circulation. The HP-ENHS is designed to preserve many features of the ENHS including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walkaway passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of advantageous features including: (1) significantly enhanced passive decay heat removal capability; (2) no positive void reactivity coefficients; (3) relatively lower corrosion of the cladding (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. This preliminary study focuses on five areas: material compatibility analysis, HP performance analysis, neutronic analysis, thermal-hydraulic analysis and safety analysis. Of the four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the

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

Generalities about nuclear reactors

International Nuclear Information System (INIS)

Jaouen, C.; Beroux, P.

2012-01-01

From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

Nuclear heat source component design considerations for HTGR process heat reactor plant concept

International Nuclear Information System (INIS)

McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

1982-01-01

Using alternate energy sources abundant in the U.S.A. to help curb foreign oil imports is vitally important from both national security and economic standpoints. Perhaps the most forwardlooking opportunity to realize national energy goals involves the integrated use of two energy sources that have an established technology base in the U.S.A., namely nuclear energy and coal. The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc.) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

Measuring device for the temperature coefficient of reactor moderators

International Nuclear Information System (INIS)

Nakano, Yuzo.

1987-01-01

Purpose: To rapidly determine by automatic calculation the temperature coefficient for moderators which has been determined so far by a log of manual processings. Constitution: Each of signals from a control rod position indicator, a reactor reactivity, instrument and moderator temperature meter are inputted, and each of the signals and designed valued for the doppler temperature coefficients are stored. Recurling calculation is conducted based on the reactivity and the moderator temperature at an interval where the temperature changes of the moderators are equalized at an identical control rod position, to determine isothermic coefficient. Then, the temperature coefficient for moderator are calculated from the isothermic coefficient and the doppler temperature coefficient. The relationship between the reactivity and the moderator temperature is plotted on a X-Y recorder. The stored signals and the calculated temperature coefficient for moderators are sequentially displayed and the results are printed out when the measurement is completed. According to the present device, since the real time processing is conducted, the processing time can be shortened remarkably. Accordingly, it is possible to save the man power for the test of the nuclear reactor and improve the reactor operation performance. (Kamimura, M.)

High-temperature-structural design and research and development for reactor system components

International Nuclear Information System (INIS)

Matsumura, Makoto; Hada, Mikio

1985-01-01

The design of reactor system components requires high-temperature-structural design guide with the consideration of the creep effect of materials related to research and development on structural design. The high-temperature-structural design guideline for the fast prototype reactor MONJU has been developed under the active leadership by Power Reactor and Nuclear Fuel Development Corporation and Toshiba has actively participated to this work with responsibility on in-vessel components, performing research and development programs. This paper reports the current status of high-temperature-structural-design-oriented research and development programs and development of analytical system including stress-evaluation program. (author)

Study on the possibility of supercritical fluid extraction for reprocessing of spent nuclear fuel from high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Duan Wuhua; Zhu Liyang; Zhu Yongjun; Xu Jingming

2011-01-01

International interest in high temperature gas-cooled reactor (HTGR) has been increasing in recent years. It is important to study on reprocessing of spent nuclear fuel from HTGR for recovery of nuclear resource and reduction of nuclear waste. Treatment of UO 2 pellets for preparing fuel elements of the 10 MW high temperature gas-cooled reactor (HTR-10) using supercritical fluid extraction was investigated. UO 2 pellets are difficult to be directly dissolved and extracted with TBP-HNO 3 complex in supercritical CO 2 (SC-CO 2 ), and the extraction efficiency is only about 7% under experimental conditions. UO 2 pellets are also difficult to be converted completely into nitrate with N 2 O 4 . When UO 2 pellets break spontaneously into U 3 O 8 powders with particle size below 100 μm under O 2 flow and 600degc, the extraction efficiency of U 3 O 8 powders with TBP-HNO 3 complex in SC-CO 2 can reach more than 98%. U 3 O 8 powders are easy to be completely converted into nitrate with N 2 O 4 . The extraction efficiency of the nitrate product with TBP in SC-CO 2 can reach more than 99%. So it has a potential prospect that application of supercritical fluid extraction in reprocessing of spent nuclear fuel from HTGR. (author)

Thermophotovoltaic Energy Conversion in Space Nuclear Reactor Power Systems

National Research Council Canada - National Science Library

Presby, Andrew L

2004-01-01

.... This has potential benefits for space nuclear reactor power systems currently in development. The primary obstacle to space operation of thermophotovoltaic devices appears to be the low heat rejection temperatures which necessitate large radiator areas...

Transient simulation of an endothermic chemical process facility coupled to a high temperature reactor: Model development and validation

International Nuclear Information System (INIS)

Brown, Nicholas R.; Seker, Volkan; Revankar, Shripad T.; Downar, Thomas J.

2012-01-01

Highlights: ► Models for PBMR and thermochemical sulfur cycle based hydrogen plant are developed. ► Models are validated against available data in literature. ► Transient in coupled reactor and hydrogen plant system is studied. ► For loss-of-heat sink accident, temperature feedback within the reactor core enables shut down of the reactor. - Abstract: A high temperature reactor (HTR) is a candidate to drive high temperature water-splitting using process heat. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Three thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. The models and coupling scheme are presented here, as well as a transient test case initiated within the chemical plant. The 50% feed flow failure within the chemical plant results in a slow loss-of-heat sink (LOHS) accident in the nuclear reactor. Due to the temperature feedback within the reactor core the nuclear reactor partially shuts down over 1500 s. Two distinct regions are identified within the coupled plant response: (1) immediate LOHS due to the loss of the sulfuric

Cooling of nuclear power stations with high temperature reactors and helium turbine cycles

International Nuclear Information System (INIS)

Foerster, S.; Hewing, G.

1977-01-01

On nuclear power stations with high temperature reactors and helium turbine cycles (HTR-single circuits) the residual heat from the energy conversion process in the primary and intermediate coolers is removed from cycled gas, helium. Water, which is circulated for safety reasons through a closed circuit, is used for cooling. The primary and intermediate coolers as well as other cooling equipment of the power plant are installed within the reactor building. The heat from the helium turbine cycle is removed to the environment most effectively by natural draught cooling towers. In this way a net plant efficiency of about 40% is attainable. The low quantities of residual heat thereby produced and the high (in comparison with power stations with steam turbine cycles) cooling agent pressure and cooling water reheat pressure in the circulating coolers enable an economically favourable design of the overall 'cold end' to be expected. In the so-called unit range it is possible to make do with one or two cooling towers. Known techniques and existing operating experience can be used for these dry cooling towers. After-heat removal reactor shutdown is effected by a separate, redundant cooling system with forced air dry coolers. The heat from the cooling process at such locations in the power station is removed to the environment either by a forced air dry cooling installation or by a wet cooling system. (orig.) [de

Kinetic characteristics of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

An, Tran Khac; Dien, Nguyen Nhi; Hien, Pham Duy [Nuclear Research Inst., Da Lat (Viet Nam); and others

1994-10-01

Kinetic characteristics of the reconstructed nuclear reactor in Dalat is investigated. Experimental parameters measured consist of: temperature coefficient of reactivity for water moderator, xenon poisoning, contribution of delayed photoneutrons induced by Be({gamma}, n) reactions and positive reactivity insertion behavior. (author). 6 refs. 4 figs.

Kinetic characteristics of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Tran Khac An; Nguyen Nhi Dien; Pham Duy Hien

1994-01-01

Kinetic characteristics of the reconstructed nuclear reactor in Dalat is investigated. Experimental parameters measured consist of: temperature coefficient of reactivity for water moderator, xenon poisoning, contribution of delayed photoneutrons induced by Be(γ, n) reactions and positive reactivity insertion behavior. (author). 6 refs. 4 figs

Gas reactor and associated nuclear experience in the UK relevant to high temperature reactor engineering

International Nuclear Information System (INIS)

Beech, D.J.; May, R.

2000-01-01

In the UK, the NNC played a leading role in the design and build of all of the UK's commercial magnox reactors and advanced gas-cooled reactors (AGRs). It was also involved in the DRAGON project and was responsible for producing designs for large scale HTRs and other gas reactor designs employing helium and carbon dioxide coolants. This paper addresses the gas reactor experience and its relevance to the current HTR designs under development which use helium as the coolant, through the consideration of a representative sample of the issues addressed in the UK by the NNC in support of the AGR and other reactor programmes. Modern HTR designs provide unique engineering challenges. The success of the AGR design, reflected in the extended lifetimes agreed upon by the licensing authorities at many stations, indicates that these challenges can be successfully overcome. The UK experience is unique and provides substantial support to future gas reactor and high temperature engineering studies. (authors)

Requirements of coolants in nuclear reactors

International Nuclear Information System (INIS)

Abass, O. A. M.

2014-11-01

This study discussed the purposes and types of coolants in nuclear reactors to generate electricity. The major systems and components associated with nuclear reactors are cooling system. There are two major cooling systems utilized to convert the heat generated in the fuel into electrical power. The primary system transfers the heat from the fuel to the steam generator, where the secondary system begins. The steam formed in the steam generator is transferred by the secondary system to the main turbine generator, where it s converted into electricity after passing through the low pressure turbine. There are various coolants used in nuclear reactors-light water, heavy water and liquid metal. The two major types of water-cooled reactors are pressurized water reactors (PWR) and boiling water reactors (BWR) but pressurized water reactors are more in the world. Also discusses this study the reactors and impact of the major nuclear accidents, in the April 1986 disaster at the Chernobyl nuclear power plant in Ukraine was the product operators, and in the March 2011 at the Fukushima nuclear power plant in Japan was the product of earthquake of magnitude 9.0, the accidents caused the largest uncontrolled radioactive release into the environment.(Author)

Nuclear reactors

International Nuclear Information System (INIS)

Prescott, R.F.; George, B.V.; Baglin, C.J.

1979-01-01

In a nuclear reactor (e.g. one having coolant down-flow through a core to a hearth below) thermal insulation (e.g. of a floor of the hearth) comprises a layer of bricks and a layer of tiles thereon, with smaller clearances between the tiles than between the bricks but with the bricks being of reduced cross-section immediately adjacent the tiles so as to be surrounded by interconnected passages, of relatively large dimensions, constituting a continuous chamber extending behind the layer of tiles. By this arrangement, lateral coolant flow in the inter-brick clearances is much reduced. The reactor core is preferably formed of hexagonal columns, supported on diamond-shaped plates each supported on a pillar resting on one of the hearth-floor tiles. Each plate has an internal duct, four upper channels connecting the duct with coolant ducts in four core columns supported by the plate, and lower channels connecting the duct to a downwardly-open recess common to three plates, grouped to form a hexagon, at their mutually-adjacent corners. This provides mixing, and temperature-averaging, of coolant from twelve columns

Progress in space nuclear reactor power systems technology development - The SP-100 program

Science.gov (United States)

Davis, H. S.

1984-01-01

Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

Proliferation resistance assessment of high temperature gas reactors

Energy Technology Data Exchange (ETDEWEB)

Chikamatsu N, M. A. [Instituto Tecnologico y de Estudios Superiores de Monterrey, Campus Santa Fe, Av. Carlos Lazo No. 100, Santa Fe, 01389 Mexico D. F. (Mexico); Puente E, F., E-mail: midori.chika@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

The Generation IV International Forum has established different objectives for the new generation of reactors to accomplish. These objectives are focused on sustain ability, safety, economics and proliferation resistance. This paper is focused on how the proliferation resistance of the High Temperature Gas Reactors (HTGR) is assessed and the advantages that these reactors present currently. In this paper, the focus will be on explaining why such reactors, HTGR, can achieve the goals established by the GIF and can present a viable option in terms of proliferation resistance, which is an issue of great importance in the field of nuclear energy generation. The reason why the HTGR are being targeted in this writing is that these reactors are versatile, and present different options from modular reactors to reactors with the same size as the ones that are being operated today. Besides their versatility, the HTGR has designed features that might improve on the overall sustain ability of the nuclear reactors. This is because the type of safety features and materials that are used open up options for industrial processes to be carried out; cogeneration for instance. There is a small section that mentions how HTGR s are being developed in the international sector in order to present the current world view in this type of technology and the further developments that are being sought. For the proliferation resistance section, the focus is on both the intrinsic and the extrinsic features of the nuclear systems. The paper presents a comparison between the features of Light Water Reactors (LWR) and the HTGR in order to be able to properly compare the most used technology today and one that is gaining international interest. (Author)

Proliferation resistance assessment of high temperature gas reactors

International Nuclear Information System (INIS)

Chikamatsu N, M. A.; Puente E, F.

2014-10-01

The Generation IV International Forum has established different objectives for the new generation of reactors to accomplish. These objectives are focused on sustain ability, safety, economics and proliferation resistance. This paper is focused on how the proliferation resistance of the High Temperature Gas Reactors (HTGR) is assessed and the advantages that these reactors present currently. In this paper, the focus will be on explaining why such reactors, HTGR, can achieve the goals established by the GIF and can present a viable option in terms of proliferation resistance, which is an issue of great importance in the field of nuclear energy generation. The reason why the HTGR are being targeted in this writing is that these reactors are versatile, and present different options from modular reactors to reactors with the same size as the ones that are being operated today. Besides their versatility, the HTGR has designed features that might improve on the overall sustain ability of the nuclear reactors. This is because the type of safety features and materials that are used open up options for industrial processes to be carried out; cogeneration for instance. There is a small section that mentions how HTGR s are being developed in the international sector in order to present the current world view in this type of technology and the further developments that are being sought. For the proliferation resistance section, the focus is on both the intrinsic and the extrinsic features of the nuclear systems. The paper presents a comparison between the features of Light Water Reactors (LWR) and the HTGR in order to be able to properly compare the most used technology today and one that is gaining international interest. (Author)

Guidelines for nuclear reactor equipments safety-analysis

International Nuclear Information System (INIS)

1978-01-01

The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others

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

Historical civilian nuclear accident based Nuclear Reactor Condition Analyzer

Science.gov (United States)

McCoy, Kaylyn Marie

There are significant challenges to successfully monitoring multiple processes within a nuclear reactor facility. The evidence for this observation can be seen in the historical civilian nuclear incidents that have occurred with similar initiating conditions and sequences of events. Because there is a current lack within the nuclear industry, with regards to the monitoring of internal sensors across multiple processes for patterns of failure, this study has developed a program that is directed at accomplishing that charge through an innovation that monitors these systems simultaneously. The inclusion of digital sensor technology within the nuclear industry has appreciably increased computer systems' capabilities to manipulate sensor signals, thus making the satisfaction of these monitoring challenges possible. One such manipulation to signal data has been explored in this study. The Nuclear Reactor Condition Analyzer (NRCA) program that has been developed for this research, with the assistance of the Nuclear Regulatory Commission's Graduate Fellowship, utilizes one-norm distance and kernel weighting equations to normalize all nuclear reactor parameters under the program's analysis. This normalization allows the program to set more consistent parameter value thresholds for a more simplified approach to analyzing the condition of the nuclear reactor under its scrutiny. The product of this research provides a means for the nuclear industry to implement a safety and monitoring program that can oversee the system parameters of a nuclear power reactor facility, like that of a nuclear power plant.

Plutonium burning in a pebble-bed type high temperature nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Bende, E.E

2000-01-24

This thesis deals with the pebble-bed High Temperature Reactor that is fuelled with pure reactor-grade plutonium. It is stressed that neither burnable poisons nor fertile materials like 238U and 212Th are present in the calculational models throughout this thesis. Chapter 2 discusses the general properties of the pebble-bed HTR: the passive safety features of this reactor; different fuel scenarios according to which the pebble-bed HTR can be operated; properties of the pebbles and the coated particles (CPs), including a concise overview of the mechanisms that can lead to coated particle failure. Special attention is paid to the effect of Pu as fuel inside these CPs thereby aiming to indicate which mechanisms are of concern when such CPs are considered as fuel in future reactors. In the last part of this chapter constraints are listed that were imposed to the models considered in the framework of this thesis. Chapter 3 presents the results of unit-cell calculations performed with three code systems. The main objective of this chapter is to compare the calculational results of one particular code system, which is a candidate for the generation of cross sections for a full-core calculation, to those of the other two code systems. Also some reactor physics interpretations of the calculational results are presented. The unit-cell calculations embrace the computation of a number of reactor physics parameters for pebbles with a varying plutonium mass per pebble and with different types of coated particles. For one pebble configuration, these parameters have been calculated for various fuel temperatures and over-all (uniform) temperatures. For that particular pebble configuration, also the results of a two burnup calculations were compared. Chapter 4 reports the results of a parameter study in which the number of coated particles per pebble as well as the type and size of the CPs have been varied. The effect of different pebble configurations on several reactor physics

Detailed Reaction Kinetics for CFD Modeling of Nuclear Fuel Pellet Coating for High Temperature Gas-Cooled Reactors

International Nuclear Information System (INIS)

Battaglia, Francine

2008-01-01

The research project was related to the Advanced Fuel Cycle Initiative and was in direct alignment with advancing knowledge in the area of Nuclear Fuel Development related to the use of TRISO fuels for high-temperature reactors. The importance of properly coating nuclear fuel pellets received a renewed interest for the safe production of nuclear power to help meet the energy requirements of the United States. High-temperature gas-cooled nuclear reactors use fuel in the form of coated uranium particles, and it is the coating process that was of importance to this project. The coating process requires four coating layers to retain radioactive fission products from escaping into the environment. The first layer consists of porous carbon and serves as a buffer layer to attenuate the fission and accommodate the fuel kernel swelling. The second (inner) layer is of pyrocarbon and provides protection from fission products and supports the third layer, which is silicon carbide. The final (outer) layer is also pyrocarbon and provides a bonding surface and protective barrier for the entire pellet. The coating procedures for the silicon carbide and the outer pyrocarbon layers require knowledge of the detailed kinetics of the reaction processes in the gas phase and at the surfaces where the particles interact with the reactor walls. The intent of this project was to acquire detailed information on the reaction kinetics for the chemical vapor deposition (CVD) of carbon and silicon carbine on uranium fuel pellets, including the location of transition state structures, evaluation of the associated activation energies, and the use of these activation energies in the prediction of reaction rate constants. After the detailed reaction kinetics were determined, the reactions were implemented and tested in a computational fluid dynamics model, MFIX. The intention was to find a reduced mechanism set to reduce the computational time for a simulation, while still providing accurate results

High-temperature reactor in modular construction

International Nuclear Information System (INIS)

Mueller, F.U.; Reutler, H.; Ullrich, M.

1981-01-01

Together with other reactors of the same type a gas-cooled, small-sized high-temperature reactor is to be assembled into a plant with modular design. The reactor vessel can be withdrawn as a whole after shutdown, removal of the fuel element charge, disassembly of the control rods, and opening of the closure of the safety containment. All apertures for the inlet and outlet of the cooling gas are located in the ground plate of the reactor. The lower part of the reactor cavern serves as inlet space for the cool gas, while the heated gas is let in through a line of a heat sink, e.g. a heat exchanger. The ground plate is connected with the hot gas line or with an inserted hot gas collecting room by means of a simple plug connection which is released automatically when the reactor vessel is withdrawn. The cooling gas, which is put into circulation by a blower and led through special conducting systems, is also used for cooling the outer metal jacket of the hot gas line. A second design is described according to which the reactor and heat exchanger are superposed in a safety containment, such as applied for pressurized water-cooled nuclear reactors. (orig.) [de

Direct digital temperature control of the A-1 nuclear reactor

International Nuclear Information System (INIS)

Karpeta, C.

1975-01-01

The application is described of one of the modern control methods for designing an experimental digital temperature control system for heavy water moderated gas cooled reactors. The synthesis of the optimal stochastic regulator for reactor control in the area of the rated steady state was carried out using the method of dynamic programming and the Kalman filter technique. The analysis of the feedback circuit was conducted using control simulation on a universal digital computer. Results and experience are summed up. (author)

Control rod drive of nuclear reactor

International Nuclear Information System (INIS)

Zhuchkov, I.I.; Gorjunov, V.S.; Zaitsev, B.I.

1980-01-01

This invention relates to nuclear reactors and, more particularly, to a drive of a control rod of a nuclear reactor and allows power control, excess reactivity compensation, and emergency shut-down of a reactor. (author)

Small high temperature gas-cooled reactors with innovative nuclear burning

International Nuclear Information System (INIS)

Liem, Peng Hong; Ismail; Sekimoto, Hiroshi

2008-01-01

Since the innovative concept of CANDLE (Constant Axial shape of Neutron Flux, nuclide densities and power shape During Life of Energy producing reactor) burning strategy was proposed, intensive research works have been continuously conducted to evaluate the feasibility and the performance of the burning strategy on both fast and thermal reactors. We learned that one potential application of the burning strategy for thermal reactors is for the High Temperature Gas-Cooled Reactors (HTGR) with prismatic/block-type fuel elements. Several characteristics of CANDLE burning strategy such as constant reactor characteristics during burn-up, no need for burn-up reactivity control mechanism, proportionality of core height with core lifetime, sub-criticality of fresh fuel elements, etc. enable us to design small sized HTGR with a high degree of safety easiness of operation and maintenance, and long core lifetime which are required for introducing the reactors into remote areas or developing countries with limited infrastructures and resources. In the present work, we report our evaluation results on small sized block-type HTGR designs with CANDLE burning strategy and compared with other existing small HTGR designs including the ones with pebble fuel elements, under both uranium and thorium fuel cycles. (author)

New advanced small and medium nuclear power reactors: possible nuclear power plants for Australia

International Nuclear Information System (INIS)

Dussol, R.J.

2003-01-01

In recent years interest has increased in small and medium sized nuclear power reactors for generating electricity and process heat. This interest has been driven by a desire to reduce capital costs, construction times and interest during construction, service remote sites and ease integration into small grids. The IAEA has recommended that the term 'small' be applied to reactors with a net electrical output less than 300 MWe and the term 'medium' to 300-700 MWe. A large amount of experience has been gained over 50 years in the design, construction and operation of small and medium nuclear power reactors. Historically, 100% of commercial reactors were in these categories in 1951-1960, reducing to 21% in 1991-2000. The technologies involved include pressurised water reactors, boiling water reactors, high temperature gas-cooled reactors, liquid metal reactors and molten salt reactors. Details will be provided of two of the most promising new designs, the South African Pebble Bed Modular Reactor (PBMR) of about 110 MWe, and the IRIS (International Reactor Innovative and Secure) reactor of about 335 MWe. Their construction costs are estimated to be about US$l,000/kWe with a generating cost for the PBMR of about US1.6c/kWh. These costs are lower than estimated for the latest designs of large reactors such as the European Pressurised Reactor (EPR) designed for 1,600 MWe for use in Europe in the next decade. It is concluded that a small or medium nuclear power reactor system built in modules to follow an increasing demand could be attractive for generating low cost electricity in many Australian states and reduce problems arising from air pollution and greenhouse gas emissions from burning fossil fuels

The experimental reactor Osiris and the nuclear fuel technology for the P.W.R. reactors

International Nuclear Information System (INIS)

Lestiboudois, G.; Contenson, G. de; Genthon, J.P.; Molvault, M.; Roche, M.

1977-01-01

The possibility of employing research reactors to study and to improve the nuclear fuel of the power reactors is presented. Measurements of temperature, pressure, stresses, thermal balance, gamma spectrometry and neutron radiography, allow the study of fuel densification, the influence of the initial filling pressure on the fission gas release and the gadolinium efficiency evolution. The solutions of the problems of failed element detection, power increase, remote handling, are presented [fr

The effects of off-center pellets on the temperature distribution and the heat flux distribution of fuel rods in nuclear reactors

International Nuclear Information System (INIS)

Peng Muzhang; Xing Jianhua

1986-01-01

This paper analyzes the effects of off-center pellets on the steady state temperature distribution and heat flux distribution of fuel rods in the nuclear reactors, and derives the dimensionless temperature distribution relationships and the dimensionless heat flux distribution relationship from the fuel rods with off-center pellets. The calculated results show that the effects of off-center will result in not only deviations of the highest temperature placement in the fuel pellets, but also the circumferentially nonuniform distributions of the temperatures and heat fluxes of the fuel rod surfaces

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1993-11-01

This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards

Tritium Formation and Mitigation in High Temperature Reactors

Energy Technology Data Exchange (ETDEWEB)

Piyush Sabharwall; Carl Stoots

2012-08-01

Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. In order to prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for the production of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750°C. Results of the diffusion model are presented for one steadystate value of tritium production in the reactor.

Sensitivity Studies of Advanced Reactors Coupled to High Temperature Electrolysis (HTE) Hydrogen Production Processes

International Nuclear Information System (INIS)

Edwin A. Harvego; Michael G. McKellar; James E. O'Brien; J. Stephen Herring

2007-01-01

High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 C to 950 C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the steam or air sweep loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycle producing the highest efficiencies varied depending on the temperature range considered

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.

Towards nuclear fusion reactors

International Nuclear Information System (INIS)

1993-11-01

The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)

Nuclear Reactor Engineering Analysis Laboratory

International Nuclear Information System (INIS)

Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

1998-01-01

The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels

Nuclear Capacity Building through Research Reactors

International Nuclear Information System (INIS)

2017-01-01

Four Instruments: •The IAEA has recently developed a specific scheme of services for Nuclear Capacity Building in support of the Member States cooperating research reactors (RR) willing to use RRs as a primary facility to develop nuclear competences as a supporting step to embark into a national nuclear programme. •The scheme is composed of four complementary instruments, each of them being targeted to specific objective and audience: Distance Training: Internet Reactor Laboratory (IRL); Basic Training: Regional Research Reactor Schools; Intermediate Training: East European Research Reactor Initiative (EERRI); Group Fellowship Course Advanced Training: International Centres based on Research Reactors (ICERR)

Artificial intelligence in nuclear reactor operation

International Nuclear Information System (INIS)

Da Ruan; Benitez-Read, J.S.

2005-01-01

Assessment of four real fuzzy control applications at the MIT research reactor in the US, the FUGEN heavy water reactor in Japan, the BR1 research reactor in Belgium, and a TRIGA Mark III reactor in Mexico will be examined through a SWOT analysis (strengths, weakness, opportunities, and threats). Special attention will be paid to the current cooperation between the Belgian Nuclear Research Centre (SCK·CEN) and the Mexican Nuclear Centre (ININ) on AI-based intelligent control for nuclear reactor operation under the partial support of the National Council for Science and Technology of Mexico (CONACYT). (authors)

Local bifurcation analysis in nuclear reactor dynamics by Sotomayor’s theorem

International Nuclear Information System (INIS)

Pirayesh, Behnam; Pazirandeh, Ali; Akbari, Monireh

2016-01-01

Highlights: • When the feedback reactivity is considered as a nonlinear function some complex behaviors may emerge in the system such as local bifurcation phenomenon. • The qualitative behaviors of a typical nuclear reactor near its equilibrium points have been studied analytically. • Comprehensive analytical bifurcation analyses presented in this paper are transcritical bifurcation, saddle- node bifurcation and pitchfork bifurcation. - Abstract: In this paper, a qualitative approach has been used to explore nuclear reactor behaviors with nonlinear feedback. First, a system of four dimensional ordinary differential equations governing the dynamics of a typical nuclear reactor is introduced. These four state variables are the relative power of the reactor, the relative concentration of delayed neutron precursors, the fuel temperature and the coolant temperature. Then, the qualitative behaviors of the dynamical system near its equilibria have been studied analytically by using local bifurcation theory and Sotomayor’s theorem. The results indicated that despite the uncertainty of the reactivity, we can analyze the qualitative behavior changes of the reactor from the bifurcation point of view. Notably, local bifurcations that were considered in this paper include transcritical bifurcation, saddle-node bifurcation and pitchfork bifurcation. The theoretical analysis showed that these three types of local bifurcations may occur in the four dimensional dynamical system. In addition, to confirm the analytical results the numerical simulations are given.

High-temperature process heat reactor with solid coolant and radiant heat exchange

International Nuclear Information System (INIS)

Alekseev, A.M.; Bulkin, Yu.M.; Vasil'ev, S.I.

1984-01-01

The high temperature graphite reactor with the solid coolant in which heat transfer is realized by radiant heat exchange is described. Neutron-physical and thermal-technological features of the reactor are considered. The reactor vessel is made of sheet carbon steel in the form of a sealed rectangular annular box. The moderator is a set of graphite blocks mounted as rows of arched laying Between the moderator rows the solid coolant annular layings made of graphite blocks with high temperature nuclear fuel in the form of coated microparticles are placed. The coolant layings are mounted onto ring movable platforms, the continuous rotation of which is realizod by special electric drives. Each part of the graphite coolant laying consecutively passes through the reactor core neutron cut-off zones and technological zone. In the core the graphite is heated up to the temperature of 1350 deg C sufficient for effective radiant heat transfer. In the neutron cut-off zone the chain reaction and further graphite heating are stopped. In the technological zone the graphite transfers the accumulated heat to the walls of technological channels in which the working medium moves. The described reactor is supposed to be used in nuclear-chemical complex for ammonia production by the method of methane steam catalytic conversion

Detectability prediction for a thermoacoustic sensor in the breazeale nuclear reactor pool

Energy Technology Data Exchange (ETDEWEB)

Smith, James [Idaho National Laboratory, Idaho Falls, ID (United States); Hrisko, Joshua [Idaho National Laboratory, Idaho Falls, ID (United States); Garrett, Steven [Idaho National Laboratory, Idaho Falls, ID (United States)

2016-03-01

Laboratory experiments have suggested that thermoacoustic engines can be in- corporated within nuclear fuel rods. Such engines would radiate sounds that could be used to measure and acoustically-telemeter information about the op- eration of the nuclear reactor (e.g., coolant temperature or uxes of neutrons or other energetic particles) or the physical condition of the nuclear fuel itself (e.g., changes in temperature, evolved gases) that are encoded as the frequency and/or amplitude of the radiated sound [IEEE Measurement and Instrumen- tation 16(3), 18-25 (2013)]. For such acoustic information to be detectable, it is important to characterize the vibroacoustical environments within reactors. Measurements will be presented of the background noise spectra (with and with- out coolant pumps) and reverberation times within the 70,000 gallon pool that cools and shields the fuel in the 1 MW research reactor on Penn State's campus using two hydrophones, a piezoelectric projector, and an accelerometer. Sev- eral signal-processing techniques will be demonstrated to enhance the measured results. Background vibrational measurement were also taken at the 250 MW Advanced Test Reactor, located at the Idaho National Laboratory, using ac- celerometers mounted outside the reactor's pressure vessel and on plumbing will also be presented. The detectability predictions made in the thesis were validated in September 2015 using a nuclear ssion-heated thermoacoustic sensor that was placed in the core of the Breazeale Nuclear Reactor on Penn State's campus. Some features of the thermoacoustic device used in that experiment will also be revealed. [Work supported by the U.S. Department of Energy.

Technology assessment HTR. Part 4. Power upscaling of High Temperature Reactors

International Nuclear Information System (INIS)

Van Heek, A.I.

1996-06-01

Designs of nuclear reactors can be classified in evolutionary, revolutionary and innovative designs. An innovative design is the High Temperature Reactor (HTR). Introduction of innovative reactors has not been successful until now. Globally, three requirements for this reactors for successful market introduction can be identified: (1) Societal support for nuclear energy, or if separable, for this reactor type, should be repaired; (2) After market introduction the innovative plant must be able to operate economically competitive; and (3) The costs of market introduction of an innovative reactor design must be limited. Until now all reactor designs classified as innovative have not yet been realized. High temperature reactors exist in many different designs. Common features are: helium coolant, graphite moderator and coated particle fuel. The combination of these creates the potential to fulfill the first requirement (public support), and similarly a hurdle to the second requirement (economical operation). All three problems existing in the eyes of the public are addressed, while a high degree of transparency is reached, making the design understandable also by others than nuclear experts. A consequence of designing according to the social support requirement is a limitation of the unit power level. The usual method to make nuclear power plants economically competitive, i.e. just raising the power level (economy of scale) could not be applied anymore. Therefore other means of cost decreasing had to be used: modularization and simplification. These ideas are explained. Since all existing HTRs are currently out of operation, additional experience from two small HTRs under construction at this moment in the Far East will be essential. In the history of HTR designs, an evolutionary path can be identified. The early designs had a philosophy of safety and economics very similar to those of LWR. Modularization was introduced to attain economic viability and the design was

Using genetic algorithms for calibrating simplified models of nuclear reactor dynamics

International Nuclear Information System (INIS)

Marseguerra, Marzio; Zio, Enrico; Canetta, Raffaele

2004-01-01

In this paper the use of genetic algorithms for the estimation of the effective parameters of a model of nuclear reactor dynamics is investigated. The calibration of the effective parameters is achieved by best fitting the model responses of the quantities of interest (e.g., reactor power, average fuel and coolant temperatures) to the actual evolution profiles, here simulated by the Quandry based reactor kinetics (Quark) code available from the Nuclear Energy Agency. Alternative schemes of single- and multi-objective optimization are investigated. The efficiency of convergence of the algorithm with respect to the different effective parameters to be calibrated is studied with reference to the physical relationships involved

Using genetic algorithms for calibrating simplified models of nuclear reactor dynamics

Energy Technology Data Exchange (ETDEWEB)

Marseguerra, Marzio E-mail: marzio.marseguerra@polimi.it; Zio, Enrico E-mail: enrico.zio@polimi.it; Canetta, Raffaele

2004-07-01

In this paper the use of genetic algorithms for the estimation of the effective parameters of a model of nuclear reactor dynamics is investigated. The calibration of the effective parameters is achieved by best fitting the model responses of the quantities of interest (e.g., reactor power, average fuel and coolant temperatures) to the actual evolution profiles, here simulated by the Quandry based reactor kinetics (Quark) code available from the Nuclear Energy Agency. Alternative schemes of single- and multi-objective optimization are investigated. The efficiency of convergence of the algorithm with respect to the different effective parameters to be calibrated is studied with reference to the physical relationships involved.

Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD

Energy Technology Data Exchange (ETDEWEB)

Moreira, Uebert G.; Dominguez, Dany S. [Universidade Estadual de Santa Cruz (UESC), Ilh´eus, BA (Brazil). Programa de P´os-Graduacao em Modelagem Computacional em Ciencia e Tecnologia; Mazaira, Leorlen Y.R.; Lira, Carlos A.B.O. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Hernandez, Carlos R.G., E-mail: uebert.gmoreira@gmail.com, E-mail: dsdominguez@gmail.com, E-mail: leored1984@gmail.com, E-mail: cabol@ufpe.br, E-mail: cgh@instec.cu [Instituto Superior de Tecnologas y Ciencias Aplicadas (InSTEC), La Habana (Cuba)

2017-07-01

The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)

Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD

International Nuclear Information System (INIS)

Moreira, Uebert G.; Dominguez, Dany S.

2017-01-01

The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)

Nuclear reactors and fuel cycle

International Nuclear Information System (INIS)

2014-01-01

The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100 th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U 3 O 8 were replaced by U 3 Si 2 -based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to fulfill its mission that is

Nuclear reactors and fuel cycle

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-07-01

The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100{sup th} nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U{sub 3}O{sub 8} were replaced by U{sub 3}Si{sub 2}-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to

Space nuclear-power reactor design based on combined neutronic and thermal-fluid analyses

International Nuclear Information System (INIS)

Koenig, D.R.; Gido, R.G.; Brandon, D.I.

1985-01-01

The design and performance analysis of a space nuclear-power system requires sophisticated analytical capabilities such as those developed during the nuclear rocket propulsion (Rover) program. In particular, optimizing the size of a space nuclear reactor for a given power level requires satisfying the conflicting requirements of nuclear criticality and heat removal. The optimization involves the determination of the coolant void (volume) fraction for which the reactor diameter is a minimum and temperature and structural limits are satisfied. A minimum exists because the critical diameter increases with increasing void fraction, whereas the reactor diameter needed to remove a specified power decreases with void fraction. The purpose of this presentation is to describe and demonstrate our analytical capability for the determination of minimum reactor size. The analysis is based on combining neutronic criticality calculations with OPTION-code thermal-fluid calculations

Automatic Control of Reactor Temperature and Power Distribution for a Daily Load following Operation

Energy Technology Data Exchange (ETDEWEB)

Yu, Keuk Jong; Kim, Han Gon [Korea Hydro and Nuclear Power Institute, Daejeon (Korea, Republic of)

2010-10-15

An automatic control method of reactor power and power distribution was developed for a daily load following operation of APR1400. This method used a model predictive control (MPC) methodology having second-order plant data. And it utilized a reactor power ratio and axial shape index as control variables. However, the reactor regulating system of APR1400 is operated by the difference between the average temperature of the reactor core and the reference temperature, which is proportional to the turbine load. Thus, this paper reports on the model predictive control methodology using fourth-order plant data and a reactor temperature instead of the reactor power shape. The purpose of this study is to develop a revised automatic controller and analyze the behavior of the nuclear reactor temperature (Tavg) and the axial shape index (ASI) using the MPC method during a daily load following operation

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)

Indian programme on molten salt cooled nuclear reactors

International Nuclear Information System (INIS)

DuIera, I.V.; Vijayan, P.K.; Sinha, R.K.

2013-01-01

Bhabha Atomic Research Centre (BARC) is developing a 600 MWth pebble bed high temperature reactor, cooled by natural circulation of molten fluoride salts and is capable of supplying process heat at 1000 ℃ to facilitate hydrogen production by splitting water. BARC has also initiated studies for a reactor concept in which salts of molten fluoride fuel and coolant in fluid form, flows through the reactor core of graphite moderator, resulting in nuclear fission within the molten salt. For thorium fuel cycle, this concept is very attractive, since the fuel can be re-processed on-line, enabling it to be an efficient neutron breeder. (author)

Heat pipe nuclear reactor for space power

Science.gov (United States)

Koening, D. R.

1976-01-01

A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

Fabrication of Tungsten-Rhenium Cladding materials via Spark Plasma Sintering for Ultra High Temperature Reactor Applications

Energy Technology Data Exchange (ETDEWEB)

Charit, Indrajit; Butt, Darryl; Frary, Megan; Carroll, Mark

2012-11-05

This research will develop an optimized, cost-effective method for producing high-purity tungsten-rhenium alloyed fuel clad forms that are crucial for the development of a very high-temperature nuclear reactor. The study will provide critical insight into the fundamental behavior (processing-microstructure- property correlations) of W-Re alloys made using this new fabrication process comprising high-energy ball milling (HEBM) and spark plasma sintering (SPS). A broader goal is to re-establish the U.S. lead in the research field of refractory alloys, such as W-Re systems, with potential applications in very high-temperature nuclear reactors. An essential long-term goal for nuclear power is to develop the capability of operating nuclear reactors at temperatures in excess of 1,000K. This capability has applications in space exploration and some special terrestrial uses where high temperatures are needed in certain chemical or reforming processes. Refractory alloys have been identified as being capable of withstanding temperatures in excess of 1,000K and are considered critical for the development of ultra hightemperature reactors. Tungsten alloys are known to possess extraordinary properties, such as excellent high-temperature capability, including the ability to resist leakage of fissile materials when used as a fuel clad. However, there are difficulties with the development of refractory alloys: 1) lack of basic experimental data on thermodynamics and mechanical and physical properties, and 2) challenges associated with processing these alloys.

Nuclear safety. Concerns about the nuclear power reactors in Cuba

International Nuclear Information System (INIS)

Wells, Jim; Aloise, Gene; Flaherty, Thomas J.; Fitzgerald, Duane; Zavala, Mario; Hayward, Mary Alice

1992-09-01

In 1976, the Soviet Union and Cuba concluded an agreement to construct two 440-megawatt nuclear power reactors near Cienfuegos on the south central coast of Cuba, about 180 miles south of Key West, Florida. The construction of these reactors, which began around 1983, was a high priority for Cuba because of its heavy dependence on imported oil. Cuba is estimated to need an electrical generation capacity of 3,000 megawatts by the end of the decade. When completed, the first reactor unit would provide a significant percentage (estimated at over 15 percent) of Cuba's need for electricity. It is uncertain when Cuba's nuclear power reactors will become operational. On September 5, 1992, Fidel Castro announced the suspension of construction at both of Cuba's reactors because Cuba could not meet the financial terms set by the Russian government to complete the reactors. Cuban officials had initially planned to start up the first of the two nuclear reactors by the end of 1993. However, before the September 5 announcement, it was estimated that this reactor would not be operational until late 1995 or early 1996. The civil construction (such as floors and walls) of the first reactor is currently estimated to be about 90 percent to 97 percent complete, but only about 37 percent of the reactor equipment (such as pipes, pumps, and motors) has been installed. The civil construction of the second reactor is about 20 percent to 30 percent complete. No information was available about the status of equipment for the second reactor. According to former Cuban nuclear power and electrical engineers and a technician, all of whom worked at the reactor site and have recently emigrated from Cuba, Cuba's nuclear power program suffers from poor construction practices and inadequate training for future reactor operators. One former official has alleged, for example, that the first reactor containment structure, which is designed to prevent the accidental release of radioactive material into

Reactors physics. Bases of nuclear physics

International Nuclear Information System (INIS)

Diop, Ch.M.

2006-01-01

The aim of nuclear reactor physics is to quantify the relevant macroscopic data for the characterization of the neutronic state of a reactor core and to evaluate the effects of radiations (neutrons and gamma radiations) on organic matter and on inorganic materials. This first article presents the bases of nuclear physics in the context of nuclear reactors: 1 - reactor physics and nuclear physics; 2 - atomic nucleus - basic definitions: nucleus constituents, dimensions and mass of the atomic nucleus, mass defect, binding energy and stability of the nucleus, strong interaction, nuclear momentums of nucleons and nucleus; 3 - nucleus stability and radioactivity: equation of evolution with time - radioactive decay law; alpha decay, stability limit of spontaneous fission, beta decay, electronic capture, gamma emission, internal conversion, radioactivity, two-body problem and notion of radioactive equilibrium. (J.S.)

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

Digital-image processing improves man-machine communication at a nuclear reactor

International Nuclear Information System (INIS)

Cook, S.A.; Harrington, T.P.; Toffer, H.

1982-01-01

The application of digital image processing to improve man-machine communication in a nuclear reactor control room is illustrated. At the Hanford N Reactor, operated by UNC Nuclear Industries for the United States Department of Energy, in Richland, Washington, digital image processing is applied to flow, temperature, and tube power data. Color displays are used to present the data in a clear and concise fashion. Specific examples are used to demonstrate the capabilities and benefits of digital image processing of reactor data. N Reactor flow and power maps for routine reactor operations and for perturbed reactor conditions are displayed. The advantages of difference mapping are demonstrated. Image processing techniques have also been applied to results of analytical reactor models; two examples are shown. The potential of combining experimental and analytical information with digital image processing to produce predictive and adaptive reactor core models is discussed. The applications demonstrate that digital image processing can provide new more effective ways for control room personnel to assess reactor status, to locate problems and explore corrective actions. 10 figures

Pellets used for nuclear reactor scram and a method for manufacturing them

International Nuclear Information System (INIS)

1974-01-01

The invention deals with a pellet to be inserted in the core of a nuclear reactor for stopping the operation of the latter. The pellet is characterized in that it is in the form of a pellet capable of rolling easily, containing a neutron poison and a solid substance undergoing a change of state when it is raised to a predetermined temperature reached by the reactor-core, that change of state causing the pellet to desintegrate and inducing the deposition of the poison. This can be applied to the shut down of gas-cooled nuclear reactors [fr

Nuclear science. U.S. electricity needs and DOE's civilian reactor development program

International Nuclear Information System (INIS)

England-Joseph, Judy; Allen, Robert E. Jr.; Fitzgerald, Duane; Young, Edward E. Jr.; Leavens, William P.; Bell, Jacqueline

1990-05-01

Electricity projections developed by the North American Electric Reliability Council (NERC) appear to be the best available estimates of future U.S. electricity needs. NERC, which represents all segments of the utility industry, forecasts that before 1998 certain regions of the country, particularly in the more heavily populated eastern half of the United States, may experience shortfalls during summer peak demand periods. These forecasts considered the utility companies' plans, as of 1989, to meet electricity needs during the period; these plans include such measures as constructing additional generators and conducting demand management programs. Working closely with the nuclear industry, DOE is supporting the development of several reactor technologies to ensure that nuclear power remains a viable electricity supply option. In fiscal year 1990, DOE's Civilian Reactor Development Program was funded at $253 million. DOE is using these funds to support industry-led efforts to develop light water reactors (LWR), advanced liquid-metal reactors (LMR), and modular high-temperature gas-cooled reactors (MHTGR) that are safe, environmentally acceptable, and economically competitive. The utility company officials we spoke with, all of whom were in the Southeast, generally supported DOE's efforts in developing these technologies. However, most of the officials do not plan to purchase nuclear reactors until after 2000 because of the high costs of constructing nuclear reactors and current public opposition to nuclear power

Hysteresis phenomenon in nuclear reactor dynamics

Energy Technology Data Exchange (ETDEWEB)

Pirayesh, Behnam; Pazirandeh, Ali [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch; Akbari, Monireh [Shahid Rajaee Teacher Training Univ., Tehran (Iran, Islamic Republic of). Dept. of Mathematics

2017-05-15

This paper applies a nonlinear analysis method to show that hysteresis phenomenon, due to the Saddle-node bifurcation, may occur in the nuclear reactor. This phenomenon may have significant effects on nuclear reactor dynamics and can even be the beginning of a nuclear reactor accident. A system of four dimensional nonlinear ordinary differential equations was considered to study the hysteresis phenomenon in a typical nuclear reactor. It should be noted that the reactivity was considered as a nonlinear function of state variables. The condition for emerging hysteresis was investigated using Routh-Hurwitz criterion and Sotomayor's theorem for saddle node bifurcation. A numerical analysis is also provided to illustrate the analytical results.

Material Control and Accounting Design Considerations for High-Temperature Gas Reactors

International Nuclear Information System (INIS)

Bjornard, Trond; Hockert, John

2011-01-01

The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC and A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC and A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC and A approaches for the two major HTGR reactor types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR (Pty) and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC and A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR and D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present

Nuclear reactor control column

International Nuclear Information System (INIS)

Bachovchin, D.M.

1982-01-01

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor

NUCLEAR REACTOR

Science.gov (United States)

Miller, H.I.; Smith, R.C.

1958-01-21

This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

Innovative designs of nuclear reactors

International Nuclear Information System (INIS)

Gabaraev, B.A.; Cherepnin, Y.S.

2010-01-01

The world development scenarios predict at least a 2.5 time increase in the global consumption of primary energy in the first half of the twenty-first century. Much of this growth can be provided by the nuclear power which possesses important advantages over other energy technologies. However, the large deployment of nuclear sources may take place only when the new generation of reactors appears on the market and will be free of the shortcomings found in the existing nuclear power installations. The public will be more inclined to accept nuclear plants that have better economics; higher safety; more efficient management of the radioactive waste; lower risk of nuclear weapons proliferation, and provided that the focus is made on the energy option free of ∇ e 2 generation. Currently, the future of nuclear power is trusted to the technology based on fast reactors and closed fuel cycle. The latter implies reprocessing of the spent nuclear fuel of the nuclear plants and re-use of plutonium produced in power reactors

The early history of high-temperature helium gas-cooled nuclear power reactors

International Nuclear Information System (INIS)

Simnad, M.T.; California Univ., San Diego, La Jolla, CA

1991-01-01

The original concepts in the proposals for high-temperature helium gas-cooled power reactors by Farrington Daniels, during the decade 1944-1955, are summarized. The early research on the development of the helium gas-cooled power reactors is reviewed, and the operational experiences with the first generation of HTGRs are discussed. (author)

Nuclear characteristic simulation device for reactor core

International Nuclear Information System (INIS)

Arakawa, Akio; Kobayashi, Yuji.

1994-01-01

In a simulation device for nuclear characteristic of a PWR type reactor, there are provided a one-dimensional reactor core dynamic characteristic model for simulating one-dimensional neutron flux distribution in the axial direction of the reactor core and average reactor power based on each of inputted signals of control rod pattern, a reactor core flow rate, reactor core pressure and reactor core inlet enthalphy, and a three-dimensional reactor core dynamic characteristic mode for simulating three-dimensional power distribution of the reactor core, and a nuclear instrumentation model for calculating read value of the nuclear instrumentation disposed in the reactor based on the average reactor core power and the reactor core three-dimensional power distribution. A one-dimensional neutron flux distribution in the axial direction of the reactor core, a reactor core average power, a reactor core three-dimensional power distribution and a nuclear instrumentation read value are calculated. As a result, the three-dimensional power distribution and the power level are continuously calculated. Further, since the transient change of the three-dimensional neutron flux distribution is calculated accurately on real time, more actual response relative to a power monitoring device of the reactor core and operation performance can be simulated. (N.H.)

The future of nuclear power and fourth-generation reactors

International Nuclear Information System (INIS)

Carre, F.; Renault, C.

2006-01-01

Faced with the exhaustion of fossil fuel resources, the output of existing nuclear power must quadruple between now and 2050, and the Commissariat a l'Energie atomique (CEA) and its industrial partners are cooperating in a programme of R and D on future nuclear power. France strategy puts rapid neutron reactors (RNR) at the forefront, in view of their possible introduction by 2040. These reactors allow a more efficient use of uranium resources and minimise the production of long-life nuclear waste. Two technologies which use respectively, sodium and gas as their coolant are being studied. For the sodium RNR, which benefits from significant existing experience, the key is to first improve its economic performance. For the gas RNR, which draws on the principles and the generic assets of the RNR, for those using helium as the coolant, and those with applications at high temperature, it is important firstly to demonstrate the key technologies such as the fuel. The decision of President Chirac to launch the study of a prototype, fourth-generation reactor for 2020 is stimulating the research effort into France future nuclear power. (author)

CFD analysis of moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor

International Nuclear Information System (INIS)

Kansal, Anuj Kumar; Joshi, Jyeshtharaj B.; Maheshwari, Naresh Kumar; Vijayan, Pallippattu Krishnan

2015-01-01

Highlights: • 3D CFD of vertical calandria vessel. • Spatial distribution of volumetric heat generation. • Effect of Archimedes number. • Non-dimensional analysis. - Abstract: Three dimensional computational fluid dynamics (CFD) analysis has been performed for the moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor under normal operating condition using OpenFOAM CFD code. OpenFOAM is validated by comparing the predicted results with the experimental data available in literature. CFD model includes the calandria vessel, calandria tubes, inlet header and outlet header. Analysis has been performed for the cases of uniform and spatial distribution of volumetric heat generation. Studies show that the maximum temperature in moderator is lower in the case of spatial distribution of heat generation as compared to that in the uniform heat generation in calandria. In addition, the effect of Archimedes number on maximum and average moderator temperature was investigated

CFD analysis of moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Kansal, Anuj Kumar, E-mail: akansal@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400094 (India); Maheshwari, Naresh Kumar, E-mail: nmahesh@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Vijayan, Pallippattu Krishnan, E-mail: vijayanp@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

2015-06-15

Highlights: • 3D CFD of vertical calandria vessel. • Spatial distribution of volumetric heat generation. • Effect of Archimedes number. • Non-dimensional analysis. - Abstract: Three dimensional computational fluid dynamics (CFD) analysis has been performed for the moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor under normal operating condition using OpenFOAM CFD code. OpenFOAM is validated by comparing the predicted results with the experimental data available in literature. CFD model includes the calandria vessel, calandria tubes, inlet header and outlet header. Analysis has been performed for the cases of uniform and spatial distribution of volumetric heat generation. Studies show that the maximum temperature in moderator is lower in the case of spatial distribution of heat generation as compared to that in the uniform heat generation in calandria. In addition, the effect of Archimedes number on maximum and average moderator temperature was investigated.

Nuclear Reactors and Technology; (USA)

Energy Technology Data Exchange (ETDEWEB)

Cason, D.L.; Hicks, S.C. (eds.)

1991-01-01

Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

Device for bonding iodine in the event of nuclear reactor accidents

International Nuclear Information System (INIS)

Hladik, O.

1988-01-01

A device for bonding iodine, in particular radioiodine released during nuclear reactor accidents, is presented. Radioiodine is bonded, even at high temperatures, so that it is neither volatile nor soluble

U.S. Nuclear Power Reactor Plant Status

Data.gov (United States)

Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

Spring unit especially intended for a nuclear reactor core

International Nuclear Information System (INIS)

Brown, S.J.; Gorholt, Wilhelm.

1977-01-01

This invention relates to a spring unit or a group of springs bearing up a sprung mass against an unsprung mass. For instance, a gas cooled high temperature nuclear reactor includes a core of relatively complex structure supported inside a casing or vessel forming a shielded cavity enclosing the reactor core. This core can be assembled from a large number of graphite blocks of different sizes and shapes joined together to form a column. The blocks of each column can be fixed together so as to form together a loose side support. Under the effect of thermal expansion and contraction, shrinkage resulting from irradiation, the effects of pressure and the contraction and creep of the reactor vessel, it is not possible to confine all the columns of the reactor core in a cylindrical rigid structure. Further, the working of the nuclear reactor requires that the reactivity monitoring components may be inserted at any time in the reactor core. A standard process consists in mounting this loosely assembled reactor core in a floating manner by keeping it away from the vessel enclosure around it by means of a number of springs fitted between the lateral surfaces of the core unit and the reactor vessel. The core may be considered as a spring supported mass whereas, relatively, the reactor vessel is a mass that is not flexibly supported [fr

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.

The fuel of nuclear reactors

International Nuclear Information System (INIS)

1995-03-01

This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

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)

Directory of Nuclear Research Reactors 1994

International Nuclear Information System (INIS)

1995-08-01

The Directory of Nuclear Research Reactors is an output of the Agency's computerized Research Reactor Data Base (RRDB). It contains administrative, technical and utilization information on research reactors known to the Agency at the end of December 1994. The data base converted from mainframe to PC is written in Clipper 5.0 and the publication generation system uses Excel 4. The information was collected by the Agency through questionnaires sent to research reactor owners. All data on research reactors, training reactors, test reactors, prototype reactors and critical assemblies are stored in the RRDB. This system contains all the information and data previously published in the Agency's publication, Directory of Nuclear Research Reactor, as well as updated information

Nuclear reactor built, being built, or planned

International Nuclear Information System (INIS)

1991-06-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1990. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE, from the US Nuclear Regulatory Commission, from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations, from US and foreign embassies, and from foreign governmental nuclear departments. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly

An endothermic chemical process facility coupled to a high temperature reactor. Part II: Transient simulation of accident scenarios within the chemical plant

International Nuclear Information System (INIS)

Brown, Nicholas R.; Revankar, Shripad T.

2012-01-01

Highlights: ► Seven quantitative transient case studies were analyzed in a coupled PBMR and thermochemical sulfur cycle based hydrogen plant. ► Positive power excursion in the nuclear reactor were found for helium-inlet overcoolings. ► In all cases studied the maximum fuel temperatures in the nuclear reactor were 200 K below the design basis limit. - Abstract: Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. Transient study of the operational or accident events within the coupled plant is largely absent from the literature. In this paper, seven quantitative transient case studies are analyzed. The case studies consist of: (1) feed flow failure from one section of the chemical plant to another with an accompanying parametric study of the temperature in an individual reaction chamber, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without emergency nuclear reactor shutdown, (6) total failure of the chemical plant, (7) control rod insertion in the nuclear reactor. Various parametric studies based on the magnitude of the events were also performed. The only chemical plant initiated events that caused a positive power excursion in the nuclear reactor were helium-inlet overcoolings due to process holding tank failures or reaction chamber ruptures. Even for a severe sustained overcooling, the calculated maximum fuel temperatures in the nuclear reactor were 200 K below the design basis limit. The qualitative basis for the case studies and the analysis models are summarized in part I of this paper.

Probabilistic method for evaluating reactivity margin of nuclear reactors

International Nuclear Information System (INIS)

Kaneko, Yoshihiko

1984-01-01

A probabilistic method is proposed that will permit in the design stage to estimate quantitatively the likelihood with which any or all design criteria applicable to a nuclear reactor are actually satisfied after its construction. The method is trially applied to the core reactivity balance problem of the experimental Very High Temperature Reactor, and calculations are performed on the probability with which a design study core will, upon construction, satisfy design criteria concerning (a) one rod stuck and (b) startup margin. The method should prove useful in making engineering judgments before approving reactor core design. (author)

Economics and utilization of thorium in nuclear reactors. Technical annexes 1 and 2

International Nuclear Information System (INIS)

1978-05-01

An assessment of the impact of utilizing the 233 U/thorium fuel cycle in the U.S. nuclear economy is strongly dependent upon several decisions involving nuclear energy policy. These decisions include: (1) to recycle or not recycle fissile material; (2) if fissile material is recycled, to recycle plutonium, 233 U, or both; and (3) to deploy or not to deploy advanced reactor designs such as Fast Breeder Reactors (FBR's), High Temperature Gas Reactors (HTGR's), and Canadian Deuterium Uranium Reactors (CANDU's). This report examines the role of thorium in the context of the above policy decisions while focusing special attention on economics and resource utilization

Temperature measurements of the aluminium claddings of fuel elements in nuclear reactor

International Nuclear Information System (INIS)

Chen Daolong

1986-01-01

A method for embedding the sheathed thermocouples in the aluminium claddings of some fuel elements of experimental reactors by ultrasonic welding technique is described. The measurement results of the cladding temperature of fuel elements in reactors are given. By means of this method, the joint between the sheathed thermocouples and the cladding of fuel elements can be made very tight, there are no bulges on the cladding surfaces, and the sheathed thermocouples are embedded strongly and reliably. Therefore an essential means is provided for acquiring the stable and dynamic state data of the cladding temperature of in-core fuel elements

Nuclear reactor

International Nuclear Information System (INIS)

Rau, P.

1980-01-01

The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

Safety of nuclear power reactors

International Nuclear Information System (INIS)

MacPherson, H.G.

1982-01-01

Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

DOE fundamentals handbook: Nuclear physics and reactor theory

International Nuclear Information System (INIS)

1993-01-01

The Nuclear Physics and Reactor Theory Handbook was developed to assist nuclear facility operating contractors in providing operators, maintenance personnel, and the technical staff with the necessary fundamentals training to ensure a basic understanding of nuclear physics and reactor theory. The handbook includes information on atomic and nuclear physics; neutron characteristics; reactor theory and nuclear parameters; and the theory of reactor operation. This information will provide personnel with a foundation for understanding the scientific principles that are associated with various DOE nuclear facility operations and maintenance

Nuclear reactor cavity floor passive heat removal system

Science.gov (United States)

Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

2018-03-06

A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluid communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.

The future of nuclear reactors

International Nuclear Information System (INIS)

Teller, E.

1989-01-01

The Atomic Energy Commission Advisory Committee on Reactor Safeguards began work in early 1948 with the firm and unanimous conviction that nuclear power could not survive a significant damaging accident. They as a committee felt that their job was to make reactors so safe that no such event would ever occur. However, ambitious reactor planners did not like all the buts and cautions that the committee was raising. They seemed to delay unduly their setting sail into the brave new world of clean, cheap, safe nuclear energy. The committee was soon nicknamed the Committee on Reactor Prevention. Reactors, of course, represented a tremendous step into the future. To an unprecedented extent, they were based on theory. But the committee did not have the luxury of putting a preliminary model into operation and waiting for difficulties to show up. In assessing new designs and developments, they had to anticipate future difficulties. Their proposals in good part were accepted, but their deep emphasis on safety did not become a part of the program. Today, forty years later, the author still believes both in the need for nuclear reactors and in the need of a thorough-going, pervasive emphasis on their safety. Real, understandable safety can be achieved, and that achievement is the key to our nuclear future. The details he gives are only examples. The need for reactors that are not only safe but obviously safe can be ignored only at our peril

Nuclear process heat at high temperature: Application, realization and development programme

International Nuclear Information System (INIS)

Sammeck, K.H.; Fischer, R.

1976-01-01

Studies in the Federal Republic of Germany (FRG), the USA and the United Kingdom have shown that high-temperature helium energy from an HTR can advantageously be utilized for coal gasification and other fossil fuel conversion processes, and that a substantial demand for substitute natural gas (SNG) can be expected in the future. These results are based on plant design studies, economic assessments and basic development efforts in the field of coal gasification with nuclear heat, which in the FRG were carried out by Arbeitsgemeinschaft Nukleare Prozesswaerme (ANP)-members, HRB and KFA Juelich. Nuclear process plants are based on different gasification processes, resulting in different concepts of the nuclear heat system. In the case of hydro-gasification it is expected that steam reformers, arranged within the primary circuit of the reactor, will be heated directly by the primary helium. In the case of steam gasification, the high-temperature energy must be transferred to the gasification process via an intermediate circuit which is coupled to a gasifier outside the containment. In both cases the design of the nuclear reactor resembles an HTR for electricity generation. The main objectives of the development of nuclear process heat are to increase the helium outlet temperature of the reactor up to 950 0 C, to develop metallic alloys for high-temperature components such as heat exchangers, to design and construct a hot-gas duct, a steam reformer and a helium-helium heat exchanger and to develop the gasification processes. The nuclear safety regulations and the interface problems between the reactor, the process plant and the electricity generating plant have to be considered thoroughly. The Arbeitsgemeinschaft Nukleare Prozesswaerme and HRB started a development programme, in close collaboration with KFA Juelich, which will lead to the construction of a prototype plant for coal gasification with nuclear heat within 5 to 5 1/2 years. A survey of the main objectives

Critical evaluation of high-temperature gas-cooled reactors applicable to coal conversion

International Nuclear Information System (INIS)

Spiewak, I.; Jones, J.E. Jr.; Rittenhouse, P.L.; DeStefano, J.R.; Delene, J.G.

1975-12-01

A critical review is presented of the technology and costs of very high-temperature gas-cooled reactors (VHTRs) applicable to nuclear coal conversion. Coal conversion processes suitable for coupling to reactors are described. Vendor concepts of the VHTR are summarized. The materials requirements as a function of process temperature in the range 1400 to 2000 0 F are analyzed. Components, environmental and safety factors, economics and nuclear fuel cycles are reviewed. It is concluded that process heat supply in the range 1400 to 1500 0 F could be developed with a high degree of assurance. Process heat at 1600 0 F would require considerably more materials development. While temperatures up to 2000 0 F appear to be attainable, considerably more research and risk were involved. A demonstration plant would be required as a step in the commercialization of the VHTR

Health requirements for nuclear reactor operators

International Nuclear Information System (INIS)

1980-05-01

The health prerequisites established for the qualification of nuclear reactor operators according to CNEN-NE-1.01 Guidelines Licensing of nuclear reactor operators, CNEN-12/79 Resolution, are described. (M.A.) [pt

Development of Very High Temperature Reactor Technology

International Nuclear Information System (INIS)

Lee, Won Jae; Noh, J. M.; Kim, Y. H.

2009-04-01

For an efficient production of nuclear hydrogen, the VHTR (Very High Temperature Gas-cooled Reactor) of 950 .deg. C outlet temperature and the interfacing system for the hydrogen production are required. We have developed various evaluation technologies for the performance and safety of VHTR through the accomplishment of this project. First, to evaluate the performance of VHTR, a series of analyses has been performed such as core characteristics at 950 .deg. C, applicability of cooled-vessel, intermediate loop system and high temperature structural integrity. Through the analyses of major accidents such as HPCC and LPCC and the analysis of the risk/performance-informed method, VHTR safety evaluation has been also performed. In addition, various design analysis codes have been developed for a nuclear design, system loop design, system performance analysis, air-ingress accident analysis, fission product/tritium transport analysis, graphite structure seismic analysis and hydrogen explosion analysis, and they are being verified and validated through a lot of international collaborations

Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

Energy Technology Data Exchange (ETDEWEB)

Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

2014-09-15

Highlights: • Gas as a coolant in Gen-IV reactors, history and development. • Main physical parameters comparison of gas coolants: carbon dioxide, helium, hydrogen with water. • Forced convection in turbulent pipe flow. • Gas cooled fast reactor concept comparisons to very high temperature reactor concept. • High temperature helium loop: concept, development, mechanism, design and constraints. - Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR). The VHTR concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behavior within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Center Rez Ltd. One of the topics analyzed in this article are also physical characteristic and benefits of gas

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.

Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels