Seawater desalination plant using nuclear heating reactor coupled with MED process

Institute of Scientific and Technical Information of China (English)

无

2000-01-01

A small size plant for seawater desalination using nuclear heating reactor coupled with MED process was developed by the Institute of Nuclear Energy Technology, Tsinghua University, China. This seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh water and energy source are severely lacking. It is also recommended as a demonstration and training facility for seawater desalination using nuclear energy. The design of small size of seawater desalination plant couples two proven technologies: Nuclear Heating Reactor (NHR) and Multi-Effect Destination (MED) process. The NHR design possesses intrinsic and passive safety features, which was demonstrated by the experiences of the project NHR-5. The intermediate circuit and steam circuit were designed as the safety barriers between the NHR reactor and MED desalination system. Within 10～200 MWt of the power range of the heating reactor, the desalination plant could provide 8000 to 150,000 m3/d of high quality potable water. The design concept and parameters, safety features and coupling scheme are presented.

Seawater desalination plant using nuclear heating reactor coupled with MED process

International Nuclear Information System (INIS)

Wu Shaorong; Dong Duo; Zhang Dafang; Wang Xiuzhen

2000-01-01

A small size plant for seawater desalination using nuclear heating reactor coupled with MED process was developed by the Institute of Nuclear Energy Technology, Tsinghua University, China. this seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh water and energy source are severely lacking. It is also recommended as a demonstration and training facility for seawater desalination using nuclear energy. The design of small size of seawater desalination plant couples two proven technologies: Nuclear Heating Reactor (NHR) and Multi-Effect Destination (MED) process. The NHR design possesses intrinsic and passive safety features, which was demonstrated by the experiences of the project NHR-5. the intermediate circuit and steam circuit were designed as the safety barriers between the NHR reactor and MED desalination system. Within 10-200 MWt of the power range of the heating reactor, the desalination plant could provide 8000 to 150,000 m 3 /d of high quality potable water. The design concept and parameters, safety features and coupling scheme are presented

Multi-stage-flash desalination plants of relative small performance with integrated pressurized water reactors as a nuclear heat source

International Nuclear Information System (INIS)

Petersen, G.; Peltzer, M.

1977-01-01

In the Krupp-GKSS joint study MINIPLEX the requirements for seawater-desalination plants with a performance in the range of 10 000 to 80 000 m 3 distillate per day heated by a nuclear reactor are investigated. The reactor concept is similar to the Integrated Pressurized Water Reactor (IPWR) of the nuclear ship OTTO HAHN. The design study shows that IPWR systems have specific advantages up to 200 MWth compared to other reactor types at least being adapted for single- and dual-purpose desalination plants. The calculated costs of the desalinated water show that due to fuel cost advantages of reactors small and medium nuclear desalination plants are economically competetive with oil-fired plants since the steep rise of oil price in autumn 1973. (author)

Small ex-core heat pipe thermionic reactor concept (SEHPTR)

International Nuclear Information System (INIS)

Jacox, M.G.; Bennett, R.G.; Lundberg, L.B.; Miller, B.G.; Drexler, R.L.

1991-01-01

The Idaho National Engineering Laboratory (INEL) has developed an innovative space nuclear power concept with unique features and significant advantages for both Defense and Civilian space missions. The Small Ex-core Heat Pipe Thermionic Reactor (SEHPTR) concept was developed in response to Air Force needs for space nuclear power in the range of 10 to 40 kilowatts. This paper describes the SEHPTR concept and discusses the key technical issues and advantages of such a system

Transient core characteristics of small molten salt reactor coupling problem between heat transfer/flow and nuclear fission reaction

International Nuclear Information System (INIS)

Yamamoto, Takahisa; Mitachi, Koshi

2004-01-01

This paper performed the transient core analysis of a small Molten Salt Reactor (MSR). The emphasis is that the numerical model employed in this paper takes into account the interaction among fuel salt flow, nuclear reaction and heat transfer. The model consists of two group diffusion equations for fast and thermal neutron fluexs, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The results of transient analysis are that (1) fission reaction (heat generation) rate significantly increases soon after step reactivity insertion, e.g., the peak of fission reaction rate achieves about 2.7 times larger than the rated power 350 MW when the reactivity of 0.15% Δk/k 0 is inserted to the rated state, and (2) the self-control performance of the small MSR effectively works under the step reactivity insertion of 0.56% Δk/k 0 , putting the fission reaction rate back on the rated state. (author)

Slowpoke: a role for nuclear technology in district heating

International Nuclear Information System (INIS)

Lynch, G.F.

1987-08-01

The successful application of the SLOWPOKE concept to satisfy the heating needs of institutions and building complexes is described. Although the load factor for heating in Japan may not be as high as those experienced in other countries of the northern hemipshere, this particular application clearly demonstrates that small, special purpose, ultra-safe nuclear energy sources are technically and economically viable. They can be designed for easy operation and maintenance, to be located in urban areas and remote communities, thereby satsifying a broad spectrum of energy needs that cannot be served by central nuclear electrical generators

Potentialities and type of integrating nuclear heating stations into district heating systems

International Nuclear Information System (INIS)

Munser, H.; Reetz, B.; Schmidt, G.

1978-01-01

Technical and economical potentialities of applying nuclear heating stations in district heating systems are discussed considering the conditions of the GDR. Special attention is paid to an optimum combination of nuclear heating stations with heat sources based on organic fuels. Optimum values of the contribution of nuclear heating stations to such combined systems and the economic power range of nuclear heating stations are estimated. Final considerations are concerned with the effect of siting and safety concepts of nuclear heating stations on the structure of the district heating system. (author)

The keys to success in marketing small heating reactors

International Nuclear Information System (INIS)

McDougall, D.S.; Lynch, G.F.

1988-01-01

The success of the SLOWPOKE Energy System requires acceptance of the SLOWPOKE reactor within the community where the reactor's energy is to be used. Public acceptance will be obtained once the public is convinced that this nuclear heat source is needed, safe and of economic benefit to the community. The need for a new application of nuclear energy is described and the ability of small reactors used for district heating to play that role is shown. The safety of the reactor is being demonstrated with the establishment of the SLOWPOKE Demonstration Reactor by Atomic Energy of Canada Limited and with open, candid discussion with the involved community. Economic arguments are reviewed and include discussion of quantitative and qualitative issues. (orig.)

Containment for low temperature district nuclear-heating reactor

International Nuclear Information System (INIS)

He Shuyan; Dong Duo

1992-03-01

Integral arrangement is adopted for Low Temperature District Nuclear-heating Reactor. Primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with reactor core. Primary coolant flows through reactor core and primary heat exchangers in natural circulation. Primary coolant pipes penetrating the wall of reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of pressure boundary of primary coolant. Therefore the small sized metallic containment closed to the wall of reactor vessel can be used for the reactor. Design principles and functions of the containment are as same as the containment for PWR. But the adoption of small sized containment brings about some benefits such as short period of manufacturing, relatively low cost, and easy for sealing. Loss of primary coolant accident would not be happened during the rupture accident of primary coolant pressure boundary inside the containment owing to its intrinsic safety

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

Low temperature nuclear heat

Energy Technology Data Exchange (ETDEWEB)

Kotakorpi, J.; Tarjanne, R. [comps.

1977-08-01

The meeting was concerned with the use of low grade nuclear heat for district heating, desalination, process heat, and agriculture and aquaculture. The sessions covered applications and demand, heat sources, and economics.

Titanium Loop Heat Pipes for Space Nuclear Radiators, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — This Small Business Innovation Research Phase I project will develop titanium Loop Heat Pipes (LHPs) that can be used in low-mass space nuclear radiators, such as...

Supply of Prague with heat from a nuclear heat source

International Nuclear Information System (INIS)

Poul, F.

1976-01-01

The proposals are discussed of supplying Prague, the Czechoslovak Capital, with nuclear reactor-generated heat energy. The proposals meet the requirements of the general urban plan of development. The first nuclear heating plant is to be sited in the Kojetice locality, in the northern Prague suburb. It will be commissioned by 1984 and 1985. It is estimated that the maximum heat output in form of hot water will be 821 MW. By 1995 the construction of the second nuclear heating plant should be started southeast or east of Prague. The connection of these two nuclear plants to the hot water mains together with other conventional heating plants will secure the heat supply for Prague and its new housing estates and industrial works. (Oy)

Market potential of small and medium-size nuclear reactors as combined heat and power plants in Europe

International Nuclear Information System (INIS)

Candeli, R.; Leuchs, D.; Garribba, S.F.; Hansen, U.; Lefevre, J.C.; Schoen, R.; Vivante, C.

1988-01-01

The preliminary results are presented concerning a study launched by the Commission of the European Communities to assess the potential market of small and medium-size nuclear reactors through EC member countries. The study was aimed at identifying those factors that may have a role in shaping the eventual deployment and diffusion of this class of nuclear reactors. In a first phase, attention focused on modular high-temperature gas-cooled reactors that would be installed to produce low-temperature heat and power. Federal Republic of Germany, Italy and France are the countries for which the investigation has been completed. The time span of interest is up to the year 2020. Referring to this horizon, an appraisal has been made of the number of nuclear units which could come on line to cope with energy demand and their timing. Through the study a distinction is made between technical potential, economic potential, and effective market potential. It is understood indeed that both economic competitiveness towards other energy sources and also institutional or organizational factors may restrict the market which could become accessible and would be covered by the new nuclear plants. (orig.)

Market potential of small and medium-size nuclear reactors as combined heat and power plants in Europe

International Nuclear Information System (INIS)

Schoen, R.; Hansen, U.; Candeli, R.; Leuchs, D.; Garribba, S.F.; Lefevre, J.C.; Vivante, C.

1987-01-01

A Study has been launched and supported by the Commission of the European Communities to assess the potential market of small and medium size nuclear reactors through EC Countries and to identify those factors which may play a dominant role in this respect. In its first phase, the Study addresses to modular high-temperature gas-cooled reactors that would be installed to produce low-temperature heat and power. Federal Republic of Germany, Italy and France are the countries for which the investigation has been either completed or is under way. The time span of interest is up to the year 2020. Referring to this horizon an appraisal has been made of the number of nuclear units which could come on line to cope with energy demand and their timing. Through the Study a distinction is made between technical potential, economic potential, and effective market potential. It is understood indeed that both economic competitiveness towards other energy sources and institutional or organizational factors may restrict the market which could be covered by this class of nuclear reactors. (author)

Experience and Prospects of Nuclear Heat Application

International Nuclear Information System (INIS)

Woite, G.; Konishi, T.; Kupitz, J.

1998-01-01

Relevant technical characteristics of nuclear reactors and heat application facilities for district heating, process heat and seawater desalination are presented and discussed. The necessity of matching the characteristics of reactors and heat applications has consequences for their technical and economic viability. The world-wide operating experience with nuclear district heating, process heating, process heat and seawater desalination is summarised and the prospects for these nuclear heat applications are discussed. (author)

Heat recovery from nuclear power plants

International Nuclear Information System (INIS)

Safa, H.

2012-01-01

The thermodynamic efficiency of a standard Nuclear Power Plant (NPP) is around 33%. Therefore, about two third of the heat generated by the nuclear fuel is literally wasted in the environment. Given the fact that the steam coming out from the high pressure turbine is superheated, it could be advantageously used for non electrical applications, particularly for district heating. Considering the technological improvements achieved these last years in heat piping insulation, it is now perfectly feasible to envisage heat transport over quite long distances, exceeding 200 km, with affordable losses. Therefore, it could be energetically wise to revise the modifications required on present reactors to perform heat extraction without impeding the NPP operation. In this paper, the case of a French reactor is studied showing that a large fraction of the wasted nuclear heat can be actually recovered and transported to be injected in the heat distribution network of a large city. Some technical and economical aspects of nuclear district heating application are also discussed. (author)

Underground nuclear power station using self-regulating heat-pipe controlled reactors

International Nuclear Information System (INIS)

Hampel, V.E.

1989-01-01

The author presents a nuclear reactor for generating electricity disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor

Underground nuclear power station using self-regulating heat-pipe controlled reactors

Science.gov (United States)

Hampel, Viktor E.

1989-01-01

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

Design and safety aspects of nuclear district heating reactors

International Nuclear Information System (INIS)

Brogli, R.; Mathews, D.; Pelloni, S.

1989-01-01

Extensive studies on the rationale, the potential and the technology of nuclear district heating have been performed in Switzerland. Beside economics the safety aspects were of primary importance. Due to the high costs to transport heat the heating reactor tend to be small and therefore, minimally staffed and located close to population centers. Stringed safety rules are therefore applying. Gas cooled reactors are well suited as district heating reactors since they have due to their characteristics several inherent features, significant safety margins and a remarkable radioactivity retention potential. Some ways to mitigate the effects of water ingress and graphite corrosion are under investigation. (author). 5 refs, 3 figs

Economic feasibility of heat supply from simple and safe nuclear plants

International Nuclear Information System (INIS)

Tian, J.

2001-01-01

Use of nuclear energy as a heating source is greatly challenged by the economic factor since the nuclear heating reactors have relative small size and often the lower plant load factor. However, use of very simple reactor could be a possible way to economically supply heat. A deep pool reactor (DPR) has been designed for this purpose. The DPR is a novel design of pool type reactor for heat only supply. The reactor core is put in a deep pool. By only putting light static water pressure on the core coolant, the DPR will be able to meet the temperature requirements of heat supply for district heating. The feature of simplicity and safety of DPR makes a decrease of investment cost compared to other reactors for heating only purposes. According to the economical assessments, the capital investment to build a DPR plant is much less than that of a pressurized reactor with pressure vessels. For the DPR with 120 or 200 MW output, it can bear the economical comparison with a usual coal-fired heating plant. Some special means taken in DPR design make an increase of the burn-up level of spent fuel and a decrease of fuel cost. The feasibility studies of DPR in some cities in China show that heating cost using nuclear energy is only one third of that by coal and only one tenth of that by nature gas. Therefore, the DPR nuclear heating system provides an economically attractive solution to satisfy the demands of district heating without contributing to increasing greenhouse gas emissions

Optimum design of a nuclear heat supply

International Nuclear Information System (INIS)

Borel, J.P.

1984-01-01

This paper presents an economic analysis for the optimum design of a nuclear heat supply to a given district-heating network. First, a general description of the system is given, which includes a nuclear power plant, a heating power plant and a district-heating network. The heating power plant is fed with steam from the nuclear power plant. It is assumed that the heating network is already in operation and that the nuclear power plant was previously designed to supply electricity. Second, a technical definition of the heat production and transportation installations is given. The optimal power of these installations is examined. The main result is a relationship between the network capacity and the level of the nuclear heat supply as a substitute for oil under the best economic conditions. The analysis also presents information for choosing the best operating mode. Finally, the heating power plant is studied in more detail from the energy, technical and economic aspects. (author)

Nuclear heat source design for an advanced HTGR process heat plant

International Nuclear Information System (INIS)

McDonald, C.F.; O'Hanlon, T.W.

1983-01-01

A high-temperature gas-cooled reactor (HTGR) coupled with a chemical process facility could produce synthetic fuels (i.e., oil, gasoline, aviation fuel, methanol, hydrogen, etc.) in the long term using low-grade carbon sources (e.g., coal, oil shale, etc.). The ultimate high-temperature capability of an advanced HTGR variant is being studied for nuclear process heat. This paper discusses a process heat plant with a 2240-MW(t) nuclear heat source, a reactor outlet temperature of 950 0 C, and a direct reforming process. The nuclear heat source outputs principally hydrogen-rich synthesis gas that can be used as a feedstock for synthetic fuel production. This paper emphasizes the design of the nuclear heat source and discusses the major components and a deployment strategy to realize an advanced HTGR process heat plant concept

Safety characteristics of small heat producing reactors

International Nuclear Information System (INIS)

Seifritz, W.

1987-10-01

The primary objectives of protection in nuclear power plants are the possibility to shut the reactor down in case of emergency and keep it subcritical in the long run, the existence of a heat sink for post-decay heat removal in order to avoid overheating, let alone core meltdown, and the containment of radioactivity within the barriers designed for this purpose, thus preventing significant activity release. In principle, these objectives can be met in various ways, namely by active, passive or inherent technical safeguards systems. In practice, a mixture of these approaches is employed in almost all cases. What matters in the end is the assessment of the overall concept, not of some outstanding feature. Inherent characteristics are easier to achieve in small reactors. However, also in this case, inherent safety does not mean absolute safety. If inherent safety characteristics were all encompassing, they would have to include self-healing effects. However, inanimate matter is incapable of such self-organization. Consequently, inherent characteristics in nuclear technology by definition should include the increased use of dissipative processes in the thermal part of the plant. (author)

Gasification with nuclear reactor heat

International Nuclear Information System (INIS)

Weisbrodt, I.A.

1977-01-01

The energy-political ultimate aims for the introduction of nuclear coal gasification and the present state of technology concerning the HTR reactor, concerning gasification and heat exchanging components are outlined. Presented on the plans a) for hydro-gasification of lignite and for steam gasification of pit coal for the production of synthetic natural gas, and b) for the introduction of a nuclear heat system. The safety and environmental problems to be expected are portrayed. The main points of development, the planned prototype plant and the schedule of the project Pototype plant Nuclear Process heat (PNP) are specified. In a market and economic viability study of nuclear coal gasification, the application potential of SNG, the possible construction programme for the FRG, as well as costs and rentability of SNG production are estimated. (GG) [de

An underground nuclear power station using self-regulating heat-pipe controlled reactors

Science.gov (United States)

Hampel, V.E.

1988-05-17

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Heat supply from nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Stach, V [Ustav Jaderneho Vyzkumu CSKAE, Rez (Czechoslovakia)

1978-05-01

The current state of world power production and consumption is assessed. Prognoses made for the years 1980 to 2000 show that nuclear energy should replace the major part of fossil fuels not only in the production of power but also in the production of heat. In this respect high-temperature reactors are highly prospective. The question is discussed of the technical and economic parameters of dual-purpose heat and power plants. It is, however, necessary to solve problems arising from the safe siting of nuclear heat and power plants and their environmental impacts. The economic benefits of combined power and heat production by such nuclear plants is evident.

Calculation of heat generation due to nuclear radiation in nuclear reactors

International Nuclear Information System (INIS)

Torres, L.M.R.; Gomes, I.C.; Maiorino, J.R.

1986-01-01

The study is performed for caculating nuclear heating due to the interaction of neutrons and gamma-rays with matter. Modifications were implemented in the ANISN code, that solves the one-dimensional transport equation using the discrete ordinate method, to include nuclear heating calculations. Tests of the implemented modifications were performed in problems of nuclear heating due to radiation energy deposition in a fusion reactor. (Author) [pt

New nuclear heat sources for district heating

International Nuclear Information System (INIS)

Lerouge, B.

1975-01-01

The means by which urban oil heating may be taken over by new energy sources, especially nuclear, are discussed. Several possibilities exist: pressurized water reactors for high powers, and low-temperature swimming-pool-type process-heat reactors for lower powers. Both these cases are discussed [fr

75 MW heat extraction from Beznau nuclear power plant (Switzerland)

International Nuclear Information System (INIS)

Handl, K.H.

1998-01-01

The district heat extraction system installed and commissioned at the Beznau Nuclear Power Plant 1983 and 1984 is working successfully since the beginning. Together with a six kilometres extension in 1994, the system now consists of a 35 kilometres main network and 85 kilometres of local distribution pipelines. The eight founding communities as well as three networks joined later have been connected. Today around 2160 consumers of the Refuna district heating, small and large private buildings, industrial and agricultural enterprises are supplied with heat from the Beznau plant (1997: 141'000 MWh). The regional district heat supply system has become an integrated part of the regional infrastructure for around 20'000 inhabitants of the lower Aare valley. Nearly 15 years of operational experience are confirming the success of the strict approval conditions for the housing connections. Remarkably deep return flow temperatures in the district heating network were leading to considerable reserves in the transport capacity of the main pipeline system. The impacts of the heat extraction from the Beznau nuclear power plant, in particular its contribution to the protection of the environment by substituting fossil fuels and preventing CO2-production, have been positive. (author)

Integral design small nuclear power plant UNITHERM

International Nuclear Information System (INIS)

Adamovich, L. A.; Grechko, G. I.; Ulasevich, V. K.; Shishkin, V. A.

1995-01-01

The need to erect expensive energy transmission lines to these places demands to use independent local energy sources. Therefore, a reasonable alternative to the plants fired fossil fuel, mostly hydrocarbon fuel, may come from the nuclear power plants (NPP) of relatively small capacity which are nonattended, shipped to the site by large-assembled modules and completely withdrawable from the site during decommissioning. Application of NPPs for power and heat supply may prove to be cost-efficient and rather positive from social and ecological point of view. UNITHERM NPP belongs to such energy sources and may be used for heat and power supply. Heat can be provided both as hot water and superheated steam. The consumers are able to specify heat/energy supply ratio. NPP design provides for independent energy supply to the consumers and the possibility to disconnect each of them without disruption of operation of the others. Thermal hydraulic diagram of UNITHERM NPP provides for the use of three interconnected, process circuits. The consumers of thermal energy (turbogenerator unit and boilers of the central heating unit) are arranged in the last circuit

Study of passive residual heat removal system of a modular small PWR reactor

International Nuclear Information System (INIS)

Araujo, Nathália N.; Su, Jian

2017-01-01

This paper presents a study on the passive residual heat removal system (PRHRS) of a small modular nuclear reactor (SMR) of 75MW. More advanced nuclear reactors, such as generation III + and IV, have passive safety systems that automatically go into action in order to prevent accidents. The purpose of the PRHRS is to transfer the decay heat from the reactor's nuclear fuel, keeping the core cooled after the plant has shut down. It starts operating in the event of fall of power supply to the nuclear station, or in the event of an unavailability of the steam generator water supply system. Removal of decay heat from the core of the reactor is accomplished by the flow of the primary refrigerant by natural circulation through heat exchangers located in a pool filled with water located above the core. The natural circulation is caused by the density gradient between the reactor core and the pool. A thermal and comparative analysis of the PRHRS was performed consisting of the resolution of the mass conservation equations, amount of movement and energy and using incompressible fluid approximations with the Boussinesq approximation. Calculations were performed with the aid of Mathematica software. A design of the heat exchanger and the cooling water tank was done so that the core of the reactor remained cooled for 72 hours using only the PRHRS

Nuclear heating - a review of projects in several countries

International Nuclear Information System (INIS)

Vymetal, L.

1980-01-01

A review is presented of projects and studies of nuclear heat generation and district heating in the USSR, France, Sweden, Finland, USA, FRG, and CSSR. Attention is primarily paid to the nuclear sources, i.e., nuclear power and heating plants and special reactors for nuclear heating plants. The questions of heat transmission and costs are also dealt with. The review is based on the literature published between 1976 and 1979. An important source were materials from the conference on the use of low-potential heat from nuclear reactors held in Otaniemi (Finland) in 1977. (author)

Hybrid district heating system with heat supply from nuclear source

International Nuclear Information System (INIS)

Havelka, Z.; Petrovsky, I.

1987-01-01

Several designs are described of heat supply from large remote power sources (e.g., WWER-1000 nuclear power plants with a 1000 MW turbine) to localities where mainly steam distribution networks have been built but only some or none networks for hot water distribution. The benefits of the designs stem from the fact that they do not require the conversion of the local steam distribution system to a hot water system. They are based on heat supply from the nuclear power plant to the consumer area in hot water of a temperature of 150 degC to 200 degC. Part of the hot water heat will be used for the production of low-pressure steam which will be compressed using heat pumps (steam compressors) to achieve the desired steam distribution network specifications. Water of lower temperature can be used in the hot water network. The hot water feeder forms an automatic pressure safety barrier in heat supply of heating or technological steam from a nuclear installation. (Z.M.). 5 figs., 9 refs

Survey of high-temperature nuclear heat application

International Nuclear Information System (INIS)

Kirch, N.; Schaefer, M.

1984-01-01

Nuclear heat application at high temperatures can be divided into two areas - use of high-temperature steam up to 550 deg. C and use of high-temperature helium up to about 950 deg. C. Techniques of high-temperature steam and heat production and application are being developed in several IAEA Member States. In all these countries the use of steam for other than electricity production is still in a project definition phase. Plans are being discussed about using steam in chemical industries, oil refineries and for new synfuel producing plants. The use of nuclear generated steam for oil recovery from sands and shale is also being considered. High-temperature nuclear process heat production gives new possibilities for the application of nuclear energy - hard coals, lignites, heavy oils, fuels with problems concerning transport, handling and pollution can be converted into gaseous or liquid energy carriers with no loss of their energy contents. The main methods for this conversion are hydrogasification with hydrogen generated by nuclear heated steam reformers and steam gasification. These techniques will allow countries with large coal resources to replace an important part of their natural gas and oil consumption. Even countries with no fossil fuels can benefit from high-temperature nuclear heat - hydrogen production by thermochemical water splitting, nuclear steel making, ammonia production and the chemical heat-pipe system are examples in this direction. (author)

THERMOS, district central heating nuclear reactors

International Nuclear Information System (INIS)

Patarin, L.

1981-02-01

In order to expand the penetration of uranium in the national energy balance sheet, the C.E.A. has been studying nuclear reactors for several years now, that are capable of providing heat at favourable economic conditions. In this paper the THERMOS model is introduced. After showing the attraction of direct town heating by nuclear energy, the author describes the THERMOS project, defines the potential market, notably in France, and applies the lay-out study to the Grenoble Nuclear Study Centre site with district communal heating in mind. The economic aspects of the scheme are briefly mentioned [fr

SECURE nuclear district heating plant

International Nuclear Information System (INIS)

Nilsson; Hannus, M.

1978-01-01

The role foreseen for the SECURE (Safe Environmentally Clean Urban REactor) nuclear district heating plant is to provide the baseload heating needs of primarily the larger and medium size urban centers that are outside the range of waste heat supply from conventional nuclear power stations. The rationale of the SECURE concept is that the simplicity in design and the inherent safety advantages due to the use of low temperatures and pressures should make such reactors economically feasible in much smaller unit sizes than nuclear power reactors and should make their urban location possible. It is felt that the present design should be safe enough to make urban underground location possible without restriction according to any criteria based on actual risk evaluation. From the environmental point of view, this is a municipal heat supply plant with negligible pollution. Waste heat is negligible, gaseous radioactivity release is negligible, and there is no liquid radwaste release. Economic comparisons show that the SECURE plant is competitive with current fossil-fueled alternatives. Expected future increase in energy raw material prices will lead to additional energy cost advantages to the SECURE plant

Heat pump augmentation of nuclear process heat

International Nuclear Information System (INIS)

Koutz, S.L.

1986-01-01

A system is described for increasing the temperature of a working fluid heated by a nuclear reactor. The system consists of: a high temperature gas cooled nuclear reactor having a core and a primary cooling loop through which a coolant is circulated so as to undergo an increase in temperature, a closed secondary loop having a working fluid therein, the cooling and secondary loops having cooperative association with an intermediate heat exchanger adapted to effect transfer of heat from the coolant to the working fluid as the working fluid passes through the intermediate heat exchanger, a heat pump connected in the secondary loop and including a turbine and a compressor through which the working fluid passes so that the working fluid undergoes an increase in temperature as it passes through the compressor, a process loop including a process chamber adapted to receive a process fluid therein, the process chamber being connected in circuit with the secondary loop so as to receive the working fluid from the compressor and transfer heat from the working fluid to the process fluid, a heat exchanger for heating the working fluid connected to the process loop for receiving heat therefrom and for transferring heat to the secondary loop prior to the working fluid passing through the compressor, the secondary loop being operative to pass the working fluid from the process chamber to the turbine so as to effect driving relation thereof, a steam generator operatively associated with the secondary loop so as to receive the working fluid from the turbine, and a steam loop having a feedwater supply and connected in circuit with the steam generator so that feedwater passing through the steam loop is heated by the steam generator, the steam loop being connected in circuit with the process chamber and adapted to pass steam to the process chamber with the process fluid

Small heating reactors for local heating of communities

International Nuclear Information System (INIS)

Seifritz, W.

1985-08-01

The incentives to introduce relatively small heating reactors for local heating of communities are presented and the reasons why this vertically integrated energy system will meet the requirement of an emission - free substitution system are outlined. (author)

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

Reactor type choice and characteristics for a small nuclear heat and electricity co-generation plant

International Nuclear Information System (INIS)

Liu Kukui; Li Manchang; Tang Chuanbao

1997-01-01

In China heat supply consumes more than 70 percent of the primary energy resource, which makes for heavy traffic and transportation and produces a lot of polluting materials such as NO x , SO x and CO 2 because of use of the fossil fuel. The utilization of nuclear power into the heat and electricity co-generation plant contributes to the global environmental protection. The basic concept of the nuclear system is an integral type reactor with three circuits. The primary circuit equipment is enclosed in and linked up directly with reactor vessel. The third circuit produces steam for heat and electricity supply. This paper presents basic requirements, reactor type choice, design characteristics, economy for a nuclear co-generation plant and its future application. The choice of the main parameters and the main technological process is the key problem of the nuclear plant design. To make this paper clearer, take for example a double-reactor plant of 450 x 2MW thermal power. There are two sorts of main technological processes. One is a water-water-steam process. Another is water-steam-steam process. Compared the two sorts, the design which adopted the water-water-steam technological process has much more advantage. The system is simplified, the operation reliability is increased, the primary pressure reduces a lot, the temperature difference between the secondary and the third circuits becomes larger, so the size and capacity of the main components will be smaller, the scale and the cost of the building will be cut down. In this design, the secondary circuit pressure is the highest among that of the three circuits. So the primary circuit radioactivity can not leak into the third circuit in case of accidents. (author)

Cogeneration using a nuclear reactor to generate process heat

International Nuclear Information System (INIS)

Alonso, Gustavo; Ramirez, Ramon

2009-01-01

Some of the new nuclear reactor technologies (Generation III+) are claiming the production of process heat as an additional value to electricity generation. These technologies are still under development and none of them has shown how this can be possible and what will be the penalty in electricity generation to have this additional product. The current study assess the likeliness of generate process heat from a Pebble Bed Modular Reactor to be used for a refinery showing different plant balance and alternatives to produce and use that process heat. An actual practical example is presented to demonstrate the cogeneration viability using the fact that the PBMR is a modular small reactor and also the challenges that this option has. (author)

Basic study for development of nuclear heat application systems

Energy Technology Data Exchange (ETDEWEB)

Inaba, Yoshitomo; Fumizawa, Motoo; Hishida, Makoto [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

1996-05-01

We need to intensely investigate real possibilities of nuclear heat application systems which exploit high potential of nuclear energy as a promising candidate of the future energy resource in the world. In this report, special interest was placed on coal reforming systems because we thought a compact heat source of nuclear power with a very high energy density might compensate the environmental problem caused by burning a great amount of coal. First, we reviewed state-of-the-art technologies for coal reforming technology with a special attention on coal gasification technologies. Based on these basic data, we proposed several nuclear coal reforming systems and discussed advantages and disadvantages of the systems. We also explored a model with which we could analyze nuclear heat application systems all together. In addition, we investigated possibility and effects of nuclear heat utilization systems producing chemical materials from carbon dioxide in flue gas of fossil fuel power plant. As a result, we showed nuclear heat application systems were useful. (author).

Model experiments on depressurisation accidents in nuclear process heat plants (HTGR)

Energy Technology Data Exchange (ETDEWEB)

Fritsching, G.; Wolf, G. [Internationale Atomreaktorbau G.m.b.H. (INTERATOM), Bergisch Gladbach (Germany, F.R.)

1981-01-15

The analysis of depressurisation accidents requires the use of digital computer programs to find out the dynamic loads acting on the plant structures. Because of the importance of such accidents in safety and licensing procedures of nuclear process heat plants, it is necessary to compare these computer results with suitable experiments to show the accuracy and the limits of the programs in question. For this purpose a series of depressurisation experiments has been started at INTERATOM on a small scale model of a primary loop of a nuclear process heat plant. Using the results of these experiments three different computer programs were tested with good success. The development of the experimental program and the estimation of the results was carried out in co-operation with KFA-Juelich and the Technische Hochschule Aachen.

Model experiments on depressurisation accidents in nuclear process heat plants (HTGR)

International Nuclear Information System (INIS)

Fritsching, G.; Wolf, G.

1981-01-01

The analysis of depressurisation accidents requires the use of digital computer programs to find out the dynamic loads acting on the plant structures. Because of the importance of such accidents in safety and licensing procedures of nuclear process heat plants, it is necessary to compare these computer results with suitable experiments to show the accuracy and the limits of the programs in question. For this purpose a series of depressurisation experiments has been started at INTERATOM on a small scale model of a primary loop of a nuclear process heat plant. Using the results of these experiments three different computer programs were tested with good success. The development of the experimental program and the estimation of the results was carried out in co-operation with KFA-Juelich and the Technische Hochschule Aachen

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

Potential of low-temperature nuclear heat applications

International Nuclear Information System (INIS)

1986-12-01

At present, more than one third of the fossil fuel currently used is being consumed to produce space heating and to meet industrial needs in many countries of the world. Imported oil still represents a large portion of this fossil fuel and despite its present relatively low price future market evolutions with consequent upward cost revisions cannot be excluded. Thus the displacement of the fossil fuel by cheaper low-temperature heat produced in nuclear power plants is a matter which deserves careful consideration. Technico-economic studies in many countries have shown that the use of nuclear heat is fully competitive with most of fossil-fuelled plants, the higher investment costs being offset by lower production cost. Another point in favour of heat generation by nuclear source is its indisputable advantage in terms of benefits to the environment. The IAEA activity plans for 1985-86 concentrate on information exchange with specific emphasis on the design criteria, operating experience, safety requirements and specifications of heat-only reactors, co-generation plants and existing power plants backfitted for additional heat applications. The information gained up to 1985 was discussed during the Advisory Group Meeting on the Potential of Low-Temperature Nuclear Heat Applications held in the Federal Institute for Reactor Research, Wuerenlingen, Switzerland in September 1985 and, is included in the present Technical Document

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

The prospects for nuclear heating in Hungary

International Nuclear Information System (INIS)

Lynch, G.F.; Papp, I.

1989-09-01

Hungary supplies only half of its energy requirements from domestic resources and is very dependent upon imports of oil, natural gas and electricity to meet the current demand. In planning to reduce the dependence on imports, nuclear technology is considered an important element in the long-term energy strategy. To this end, an aggressive nuclear electricity generation program is being implemented with four 440 MWe units now operating and two 1000 MWe units committed. However, nuclear technology must be used in other energy sectors if the goal of long-term energy independence is to be achieved. On the demand side, 30% of the primary energy is consumed in the public sector, the major component being residential heating. Of the 3.7 million apartments in Hungary, 500 000 benefit from being connected to municipal district heating systems that use natural gas or oil as the energy base. This is, therefore, another significant energy sector that is amenable to using nuclear technology to substitute for imported oil and natural gas. In assessing alternative nuclear heat sources, a joint study was undertaken between Canada and Hungary to determine the feasibility of using the SLOWPOKE Energy System that has recently been developed. The SLOWPOKE Energy System is a benign nuclear heat source designed to supply 10 thermal megawatts in the form of hot water for local heating systems in buildings and institutions. It uses a combination of inherent safety features, including natural convection circulation and negative reactivity coefficients, and engineered features to ensure an extremely safe system. A SLOWPOKE demonstration heating reactor has been constructed in Canada. The unit started operation in 1987 July and is currently undergoing an extensive test program

Survey of heat-pipe application under nuclear environment

International Nuclear Information System (INIS)

Tsuyuzaki, Noriyoshi; Saito, Takashi; Okamoto, Yoshizo; Hishida, Makoto; Negishi, Kanji.

1986-11-01

Heat pipes today are employed in a wide variety of special heat transfer applications including nuclear reactor. In this nuclear technology area in Japan, A headway speed of the heat pipe application technique is not so high because of safety confirmation and investigation under each developing step. Especially, the outline of space craft is a tendency to increase the size. Therefore, the power supply is also tendency to increase the outlet power and keep the long life. Under SP-100 project, the development of nuclear power supply system which power is 1400 - 1600 KW thermal and 100 KW electric power is steadily in progress. Many heat pipes are adopted for thermionic conversion and coolant system in order to construct more safety and light weight system for the project. This paper describes the survey of the heat pipe applications under the present and future condition for nuclear environment. (author)

Optimization of heat supply systems employing nuclear power plants

International Nuclear Information System (INIS)

Urbanek, J.

1988-01-01

Decision making on the further development of heat supply systems requires optimization of the parameters. In particular, meeting the demands of peak load ranges is of importance. The heat supply coefficient α and the annual utilization of peak load equipment τ FS have been chosen as the characteristic quantities to describe them. The heat price at the consumer, C V , offers as the optimization criterion. The transport distance, temperature spread of the heating water, and different curves of annual variation of heat consumption on heat supply coefficient and heat price at the consumer. A comparison between heat supply by nuclear power plants and nuclear heating stations verifies the advantage of combined heat and power generation even with longer heat transport distances as compared with local heat supply by nuclear district heating stations based on the criterion of minimum employment of peak load boilers. (author)

High temperature nuclear heat for isothermal reformer

International Nuclear Information System (INIS)

Epstein, M.

2000-01-01

High temperature nuclear heat can be used to operate a reformer with various feedstock materials. The product synthesis gas can be used not only as a source for hydrogen and as a feedstock for many essential chemical industries, such as ammonia and other products, but also for methanol and synthetic fuels. It can also be burnt directly in a combustion chamber of a gas turbine in an efficient combined cycle and generate electricity. In addition, it can be used as fuel for fuel cells. The reforming reaction is endothermic and the contribution of the nuclear energy to the calorific value of the final product (synthesis gas) is about 25%, compared to the calorific value of the feedstock reactants. If the feedstock is from fossil origin, the nuclear energy contributes to a substantial reduction in CO 2 emission to the atmosphere. The catalytic steam reforming of natural gas is the most common process. However, other feedstock materials, such as biogas, landfill gas and CO 2 -contaminated natural gas, can be reformed as well, either directly or with the addition of steam. The industrial steam reformers are generally fixed bed reactors, and their performance is strongly affected by the heat transfer from the furnace to the catalyst tubes. In top-fired as well as side-fired industrial configurations of steam reformers, the radiation is the main mechanism of heat transfer and convection heat transfer is negligible. The flames and the furnace gas constitute the main sources of the heat. In the nuclear reformers developed primarily in Germany, in connection with the EVA-ADAM project (closed cycle), the nuclear heat is transferred from the nuclear reactor coolant gas by convection, using a heating jacket around the reformer tubes. In this presentation it is proposed that the helium in a secondary loop, used to cool the nuclear reactor, will be employed to evaporate intermediate medium, such as sodium, zinc and aluminum chloride. Then, the vapors of the medium material transfer

Safety study on nuclear heat utilization system - accident delineation and assessment on nuclear steelmaking pilot plant

International Nuclear Information System (INIS)

Yoshida, T.; Mizuno, M.; Tsuruoka, K.

1982-01-01

This paper presents accident delineation and assessment on a nuclear steelmaking pilot plant as an example of nuclear heat utilization systems. The reactor thermal energy from VHTR is transported to externally located chemical process plant employing helium-heated steam reformer by an intermediate heat transport loop. This paper on the nuclear steelmaking pilot plant will describe (1) system transients under accident conditions, (2) impact of explosion and fire on the nuclear reactor and the public and (3) radiation exposure on the public. The results presented in this paper will contribute considerably to understanding safety features of nuclear heat utilization system that employs the intermediate heat transport loop and the helium-heated steam reformer

District heating grid of the Daqing Nuclear Heating Plant

Energy Technology Data Exchange (ETDEWEB)

Changwen, Ma [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

1997-09-01

The Daqing Nuclear Heating Plant is the first commercial heating plant to be built in China. The plant is planned to be used as the main heat resource of one residential quarter of Daqing city. The main parameters of the heating plant are summarized in the paper. The load curve shows that the capacity of the NHP is about 69% of total capacity of the grid. The 12 existing boilers can be used as reserve and peak load heat resources. Two patterns of load following have have been considered and tested on the 5MW Test Heating Reactor. Experiment shows load of heat grid is changed slowly, so automatic load following is not necessary. (author). 9 figs, 1 tab.

French nuclear power plants for heat generation

International Nuclear Information System (INIS)

Girard, Y.

1984-01-01

The considerable importance that France attributes to nuclear energy is well known even though as a result of the economic crisis and the energy savings it is possible to observe a certain downward trend in the rate at which new power plants are being started up. In July 1983, a symbolic turning-point was reached - at more than 10 thousand million kW.h nuclear power accounted, for the first time, for more than 50% of the total amount of electricity generated, or approx. 80% of the total electricity output of thermal origin. On the other hand, the direct contribution - excluding the use of electricity - of nuclear energy to the heat market in France remains virtually nil. The first part of this paper discusses the prospects and realities of the application, at low and intermediate temperatures, of nuclear heat in France, while the second part describes the French nuclear projects best suited to the heat market (excluding high temperatures). (author)

Modeling transient heat transfer in nuclear waste repositories.

Science.gov (United States)

Yang, Shaw-Yang; Yeh, Hund-Der

2009-09-30

The heat of high-level nuclear waste may be generated and released from a canister at final disposal sites. The waste heat may affect the engineering properties of waste canisters, buffers, and backfill material in the emplacement tunnel and the host rock. This study addresses the problem of the heat generated from the waste canister and analyzes the heat distribution between the buffer and the host rock, which is considered as a radial two-layer heat flux problem. A conceptual model is first constructed for the heat conduction in a nuclear waste repository and then mathematical equations are formulated for modeling heat flow distribution at repository sites. The Laplace transforms are employed to develop a solution for the temperature distributions in the buffer and the host rock in the Laplace domain, which is numerically inverted to the time-domain solution using the modified Crump method. The transient temperature distributions for both the single- and multi-borehole cases are simulated in the hypothetical geological repositories of nuclear waste. The results show that the temperature distributions in the thermal field are significantly affected by the decay heat of the waste canister, the thermal properties of the buffer and the host rock, the disposal spacing, and the thickness of the host rock at a nuclear waste repository.

Promising design options for the encapsulated nuclear heat source reactor

International Nuclear Information System (INIS)

Conway, L.; Carelli, M.D.; Dzodzo, M.; Hossain, Q.; Brown, N.W.; Wade, D.C.; Sienick, J.J.; Greenspan, E.; Kastenberg, W.E.; Saphier, D.

2001-01-01

Promising design options for the Encapsulated Nuclear Heat Source (ENHS) liquid-metal cooled fast reactor were identified during the first year of the DOE NERI program sponsored feasibility study. Many opportunities for incorporation of innovations in design and fabrication were identified. Three of the innovations are hereby described: a novel IHX (intermediate heat exchanger) made of a relatively small number of rectangular channels, an ENHS module design featuring 100% natural circulation, and a novel conceptual design of core support and fuelling. As a result of the first year study the ENHS concept appears more practical and more promising than perceived at the outset of this study. (authors)

Promising design options for the encapsulated nuclear heat source reactor

Energy Technology Data Exchange (ETDEWEB)

Conway, L.; Carelli, M.D.; Dzodzo, M. [Westinghouse Science and Technology, Pittsburgh, PA (United States); Hossain, Q.; Brown, N.W. [Lawrence Livermore National Lab., CA (United States); Wade, D.C.; Sienick, J.J. [Argonne National Lab., IL (United States); Greenspan, E.; Kastenberg, W.E.; Saphier, D. [University of California Dept of Nuclear Engineering, Berkeley, CA (United States)

2001-07-01

Promising design options for the Encapsulated Nuclear Heat Source (ENHS) liquid-metal cooled fast reactor were identified during the first year of the DOE NERI program sponsored feasibility study. Many opportunities for incorporation of innovations in design and fabrication were identified. Three of the innovations are hereby described: a novel IHX (intermediate heat exchanger) made of a relatively small number of rectangular channels, an ENHS module design featuring 100% natural circulation, and a novel conceptual design of core support and fuelling. As a result of the first year study the ENHS concept appears more practical and more promising than perceived at the outset of this study. (authors)

Economics of long distance transmission, storage and distribution of heat from nuclear plants with existing and newer techniques

International Nuclear Information System (INIS)

Margen, Peter

1977-01-01

Nuclear plants can provide heat for district heating in mainly two ways. Central nuclear power plants sufficiently large to be economic as electricity producers could instead be designed for heat extraction at temperatures useful for district heating. The second promising way is to design simple low temperature reactors, so simple and safe that near urban location becomes feasible. The manner of transport distribution and storage of heat is discussed in this paper which are very important especially in the cost calculations. The economic objectives can often be attained already with conventional technigues even when transport distances are large. But newer techniques of transport promise to make even cities at greater distances from major nuclear power plants economically connectible whilst new techniques for small distribution pipes help to extend the economic distribution area to the less dense one-family house districts. (M.S.)

Prospects for the utilization of small nuclear plants for civil ships, floating heat and power stations and power seawater desalination complexes

International Nuclear Information System (INIS)

Polunichev, V.I.

2000-01-01

Small power nuclear reactor plants developed by OKB Mechanical Engineering are widely used as propulsion plants in various civil ships. Russia is the sole country in the world that possesses a powerful icebreaker and transport fleet which offers effective solution for vital socio-economic tasks of Russia's northern regions by maintaining a year-round navigation along the Arctic sea route. In the future, intensification of freighting volumes is expected in Arctic seas and at estuaries of northern rivers. Therefore, further replenishment of nuclear-powered fleet is needed by new generation ice-breakers equipped with advanced reactor plants. Adopted progressive design and technology solutions, reliable equipment and safety systems being continuously perfected on the basis of multi year operation experience feedback, addressing updated safety codes and achievement of science and technology, allow the advanced propulsion reactor plants of this type to be recommended as energy sources for floating heat and power co-generation stations and power-seawater desalination complexes. (author)

Nuclear district heating

International Nuclear Information System (INIS)

Ricateau, P.

1976-01-01

An economic study of nuclear district heating is concerned with: heat production, its transmission towards the area to be served and the distribution management towards the consumers. Foreign and French assessments show that the high cost of now existing techniques of hot water transport defines the competing limit distance between the site and township to be below some fifty kilometers for the most important townships (provided that the fuel price remain stationary). All studies converge towards the choice of a high transport temperature as soon as the distance is of some twenty kilometers. As for fossile energy saving, some new possibilities appear with process heat reactors; either PWR of about 1000MWth for large townships, or pool-type reactors of about 100MWth when a combination with an industrial steam supply occurs [fr

Development and construction of nuclear power and nuclear heating stations in the USSR

International Nuclear Information System (INIS)

Schmidt, G.; Kirmse, B.

1983-01-01

The state-of-the-art of nuclear power technology in the USSR is reviewed by presenting characteristic data on design and construction. The review takes into consideration the following types of facilities: Nuclear power stations with 1000 MWe pressurized water reactors, with 1000 MWe pressure tube boiling water reactors, and with 600 MWe fast breeder reactors; nuclear heating power stations with 1000 MWe reactors and nuclear heating stations with 500 MWth boiling water reactors

Analysis on flow characteristic of nuclear heating reactor

International Nuclear Information System (INIS)

Jiang Shengyao; Wu Xinxin

1997-06-01

The experiment was carried out on the test loop HRTL-5, which simulates the geometry and system design of a 5 MW Nuclear heating reactor. The analysis was based on a one-dimensional two-phase flow drift model with conservation equations for mass, steam mass, energy and momentum. Clausius-Clapeyron equation was used for the calculation of flashing front in the riser. A set of ordinary equation, which describes the behavior of two-phase flow in the natural circulation system, was derived through integration of the above conservation equations in subcooled boiling region, bulk boiling region in the heated section and in the riser. The method of time-domain was used for the calculation. Both static and dynamic results are presented. System pressure, inlet subcooling and heat flux are varied as input parameters. The results show that, firstly, subcooled boiling in the heated section and void flashing in the riser have significant influence on the distribution of the void fraction, mass flow rate and stability of the system, especially at lower pressure, secondly, in a wide range of two-phase flow conditions, only subcooled boiling occurs in the heated section. For the designed two-phase regime operation of the 5 MW nuclear heating reactor, the temperature at the core exit has not reaches its saturation value. Thirdly, the mechanism of two-phase flow oscillation, namely, 'zero-pressure-drop', is described. In the wide range of inlet subcooling (0 K<ΔT<28 K) there exists three regions for system flow condition, namely, (1) stable two-phase flow, (2) bulk and subcooled boiling unstable flow, (3) subcooled boiling and single phase stable flow. The response of mass flow rate, after a small disturbance in the heat flux, is showed in the above inlet subcooling range, and based on it the instability map of the system is given through experiment and calculation. (3 refs., 9 figs.)

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

Heat extraction from turbines of Czechoslovak nuclear power plants for district heating

International Nuclear Information System (INIS)

Drahy, J.

1985-01-01

Two design are described of SKODA extraction turbines for Czechoslovak nuclear power plants with WWER-440 and WWER-1000 reactors. 220 MW steam turbines were originally designed as pure condensation turbines with uncontrolled steam extraction. Optimal ways are now being sought for their use for heating hot water for district heating. For district heating of the town of Trnava, the nuclear power plant at Jaslovske Bohunice will provide a two-step heating of water from 70 to 120 degC with a heat supply of 60 MW th from one turbine unit. The ratio of obtained heat power to lost electric power is 5.08. Investigations showed the possibility of extracting 85 MW th of heat from uncontrolled steam extraction, this at three-step water heating from 60 to 145 degC, the ratio of gained and lost power being 7.14. Information is presented on the SKODA 220 MW turbine with steam extraction for heat supply purposes and on the 1000 MW turbine with 893 MW th heat extraction. The specifications of both types are given. (Pu)

Annual harvests of Corbicula populations prevent clogging of nuclear reactor heat exchangers

International Nuclear Information System (INIS)

Harvey, R.S.

1983-01-01

An annual program for removal of millions of Corbicula from upstream cooling water basins has prevented reclogging of nuclear reactor heat exchanger distributor plates at the Savannah River Plant during the past seven years. There are nine 32-megaliter basins in the three operating reactor areas where some settling of particulates occurs before cooling water is passed through screens in route to heat exchangers. Annual cleanings keep silt/clam substrate levels low and clam sizes small. Data are presented on the size/age distribution for clams recolonizing basins between cleanings

Activities in the field of small nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Baranaev, Yu.D.; Dolgov, V.V.; Sergeev, Yu.A. [Physics and Power Eng. Inst., Obninsk (Russian Federation). State Res. Centre

1997-10-01

Considerable efforts have been undertaken for development, design, construction and operation of small nuclear power plants (SNPP) in Russia. Systematic work in this area was started in the mid-1950s. The driving force for this activity was the awareness that the use of nuclear fuel would practically solve the problem of fuel transportation. As far as the remote northern regions are concerned, this provides the key advantage of nuclear over conventional energy sources. The activity in the field of SNPP has included pre-design analytical feasibility studies and experimental research including large-scale experiments on critical assemblies, thermal and hydraulic test facilities, research and development work, construction and operation of pilot and demonstration SNPPs, and finally, construction and more than 20 years of operation of the commercial SNPP, namely Bilibino nuclear co-generation plant (NCGP) located in Chukotka autonomous district, which is one of the most remote regions in the far north-east of Russia. In recent years, studies have been carried out on the development of several new SNPP designs using advanced reactors of the new generation. Among these are the second stage of Bilibino NCGP, floating NCGP VOLNOLOM-3, designated for siting in the Arctic sea coast area, and a nuclear district heating plant for the town of Apatity, in the Murmansk region. In this paper, the background and current status of the SNPPs are given, and the problems as well as prospects of small nuclear reactors development and implementation are considered. (orig.) 20 refs.

Chinese nuclear heating test reactor and demonstration plant

International Nuclear Information System (INIS)

Wang Dazhong; Ma Changwen; Dong Duo; Lin Jiagui

1992-01-01

In this report the importance of nuclear district heating is discussed. From the viewpoint of environmental protection, uses of energy resources and transport, the development of nuclear heating in China is necessary. The development program of district nuclear heating in China is given in the report. At the time being, commissioning of the 5 MW Test Heating Reactor is going on. A 200 MWt Demonstration Plant will be built. In this report, the main characteristics of these reactors are given. It shows this type of reactor has a high inherent safety. Further the report points out that for this type of reactor the stability is very important. Some experimental results of the driving facility are included in the report. (orig.)

Nuclear reactor auxiliary heat removal system

International Nuclear Information System (INIS)

Thompson, R.E.; Pierce, B.L.

1977-01-01

An auxiliary heat removal system to remove residual heat from gas-cooled nuclear reactors is described. The reactor coolant is expanded through a turbine, cooled in a heat exchanger and compressed by a compressor before reentering the reactor coolant. The turbine powers both the compressor and the pump which pumps a second fluid through the heat exchanger to cool the reactor coolant. A pneumatic starter is utilized to start the turbine, thereby making the auxiliary heat removal system independent of external power sources

Nuclear fuel safety studies by laser pulse heating

International Nuclear Information System (INIS)

Viswanadham, C.S.; Kumar, Santosh; Dey, G.K.; Kutty, T.R.G.; Khan, K.B.; Kumar, Arun; Jathar, V.P.; Sahoo, K.C.

2009-01-01

The behaviour of nuclear fuels under transient heating conditions is vital to nuclear safety. A laser pulse based heating system to simulate the transient heating conditions experienced by the fuel during reactor accidents like LOCA and RIA is under development at BARC, Mumbai. Some of the concepts used in this system are under testing in pilot studies. This paper describes the results of some pilot studies carried out on unirradiated UO 2 specimens by laser pulse heating, followed by metallography and X-ray diffraction measurements. (author)

Status of non-electric nuclear heat applications: Technology and safety

International Nuclear Information System (INIS)

2000-11-01

Nuclear energy plays an important role in electricity generation, producing 16% of the world's electricity at the beginning of 1999. It has proven to be safe, reliable, economical and has only a minimal impact on the environment. Most of the world's energy consumption, however, is in the form of heat. The market potential for nuclear heat was recognized early. Some of the first reactors were used for heat supply, e.g. Calder Hall (United Kingdom), Obninsk (Russian Federation), and Agesta (Sweden). Now, over 60 reactors are supplying heat for district heating, industrial processes and seawater desalination. But the nuclear option could be better deployed if it would provide a larger share of the heat market. In particular, seawater desalination using nuclear heat is of increasing interest to some IAEA Member States. In consideration of the growing experience being accumulated, the IAEA periodically reviews the progress and new developments in the field of nuclear heat applications. This publication summarizes the recent activities among Member States presented at a Technical Committee meeting in April 1999. The purpose of the meeting was to provide a forum for the exchange of up to date information on the prospect, design, safety and licensing aspects, and development of non-electrical applications of nuclear heat for industrial use. This mainly included seawater desalination and hydrogen production

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.

Heat transfer and fluid flow in nuclear systems

CERN Document Server

Fenech, Henri

1982-01-01

Heat Transfer and Fluid in Flow Nuclear Systems discusses topics that bridge the gap between the fundamental principles and the designed practices. The book is comprised of six chapters that cover analysis of the predicting thermal-hydraulics performance of large nuclear reactors and associated heat-exchangers or steam generators of various nuclear systems. Chapter 1 tackles the general considerations on thermal design and performance requirements of nuclear reactor cores. The second chapter deals with pressurized subcooled light water systems, and the third chapter covers boiling water reacto

Materials for nuclear diffusion-bonded compact heat exchangers

International Nuclear Information System (INIS)

Li, Xiuqing; Smith, Tim; Kininmont, David; Dewson, Stephen John

2009-01-01

This paper discusses the characteristics of materials used in the manufacture of diffusion bonded compact heat exchangers. Heatric have successfully developed a wide range of alloys tailored to meet process and customer requirements. This paper will focus on two materials of interest to the nuclear industry: dual certified SS316/316L stainless steel and nickel-based alloy Inconel 617. Dual certified SS316/316L is the alloy used most widely in the manufacture of Heatric's compact heat exchangers. Its excellent mechanical and corrosion resistance properties make it a good choice for use with many heat transfer media, including water, carbon dioxide, liquid sodium, and helium. As part of Heatric's continuing product development programme, work has been done to investigate strengthening mechanisms of the alloy; this paper will focus in particular on the effects of nitrogen addition. Another area of Heatric's programme is Alloy 617. This alloy has recently been developed for diffusion bonded compact heat exchanger for high temperature nuclear applications, such as the intermediate heat exchanger (IHX) for the very high temperature nuclear reactors for production of electricity, hydrogen and process heat. This paper will focus on the effects of diffusion bonding process and cooling rate on the properties of alloy 617. This paper also compares the properties and discusses the applications of these two alloys to compact heat exchangers for various nuclear processes. (author)

Trends in safety objectives for nuclear district heating plants

Energy Technology Data Exchange (ETDEWEB)

Brogli, R [Paul Scherrer Inst., Villigen (Switzerland)

1997-09-01

Safety objectives for dedicated nuclear heating plants are strongly influenced on the one hand by what is accepted for electricity nuclear stations, and on the other hand by the requirement that for economical reasons heating reactors have to be located close to population centers. The paper discusses the related trends and comes to the conclusion that on account of the specific technical characteristics of nuclear heating plants adequate safety can be provided even for highly populated sites. (author). 8 refs.

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

High temperature nuclear process heat systems for chemical processes

International Nuclear Information System (INIS)

Jiacoletti, R.J.

1976-01-01

The development planning and status of the very high temperature gas cooled reactor as a source of industrial process heat is presented. The dwindling domestic reserves of petroleum and natural gas dictate major increases in the utilization of coal and nuclear sources to meet the national energy demand. The nuclear process heat system offers a unique combination of the two that is environmentally and economically attractive and technically sound. Conceptual studies of several energy-intensive processes coupled to a nuclear heat source are presented

Development of dedicated nuclear heating plants in the Federal Republic of Germany

International Nuclear Information System (INIS)

Goetzmann, C.A.; Schroeter, K.E.

1987-01-01

In many cases district heating is both economically and environmentally superior over directly burning of fossil fuels in individual furnaces primarily because of the efficiency of the usual applied co-generation principle with regard to fuel utilization and flue gas clean up. In principle, this argument should carry even greater weight in conjunction with nuclear energy. The major draw back of dedicated heating reactors as seen up to now concerns the high specific capital cost for the application-dictated small reactor size. The paper discusses by way of a specific example in what directions solutions are being sought. (author)

Development of dedicated nuclear heating plants in the Federal Republic of Germany

International Nuclear Information System (INIS)

Goetzmann, C.A.; Schroeter, K.E.

1988-01-01

In many cases district heating is both economically and environmentally superior over directly burning of fossil fuels in individual furnaces primarily because of the efficiency of the usual applied co-generation principle with regard to fuel utilization and flue gas clean up. In principle, this argument should carry even greater weight in conjunction with nuclear energy. The major draw back of dedicated heating reactors as seen up to now concerns the high specific capital cost for the application-dictated small reactor size. The paper discusses by way of a specific example in what directions solutions are being sought

Improving the fidelity of electrically heated nuclear systems testing using simulated neutronic feedback

International Nuclear Information System (INIS)

Bragg-Sitton, Shannon M.; Godfroy, Thomas J.; Webster, Kenny

2010-01-01

Nonnuclear test platforms and methodologies can be employed to reduce the overall cost, risk and complexity of testing nuclear systems while allowing one to evaluate the operation of an integrated nuclear system within a reasonable timeframe, providing valuable input to the overall system design. In a nonnuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard electric test techniques allow one to fully assess thermal, heat transfer, and stress related attributes of a given system, but these approaches fail to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and testing with nuclear fuel elements installed. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. This paper summarizes the results of initial system dynamic response testing for two electrically heated reactor concepts: a heat pipe-cooled reactor simulator with integrated heat exchanger and a gas-cooled reactor simulator with integrated Brayton power conversion system. Initial applications apply a simplified reactor kinetics model with either a single or an averaged measured state point. Preliminary results demonstrate the applicability of the dynamic test methodology to any reactor type, elucidating the variation in system response characteristics in different reactor concepts. These results suggest a need to further enhance the dynamic test approach by incorporating a more accurate model of the reactor dynamics and improved hardware instrumentation for better state estimation in application of the

Research and development of the Chinese nuclear heating reactor

Energy Technology Data Exchange (ETDEWEB)

Dazhong, Wang; Wenziang, Zheng; Jiangui, Lin; Changwen, Ma; Duo, Dong [Institute of Nuclear Energy and Technology, Tsinghua Univ., Beijing (China)

1997-09-01

The paper presents the significance of nuclear heat application in China as well as the development status, main design features and safety concepts of the nuclear heating reactor exploited by INET. (author). 3 refs, 3 figs, 1 tab.

A heated large block test for high level nuclear waste management

International Nuclear Information System (INIS)

Lin, W.; Wilder, D.G.; Blink, J.A.; Blair, S.C.; Buscheck, T.A.; Glassley, W.E.; Lee, K.; Owens, M.W.; Roberts, J.J.

1995-01-01

The radioactive decay heat from high-level nuclear waste may, depending on the thermal load, create coupled thermal-mechanical-hydrological-chemical (TMHC) processes in the host rock of a repository. A heated large block test (LBT) is designed to understand some of the TNMC processes. A block of Topopah Spring tuff of about 3 x 3 x 4.5 m was isolated at Fran Ridge, Nevada Test Site. Small blocks of the rock adjacent to the large block were collected for laboratory testing of some individual thermal-mechanical, thermal-hydrological, and thermal-chemical processes. The large block will be heated by heaters within so that a dryout zone and a condensate zone will exist simultaneously. Guard heaters on the block sides will be used to minimize horizontal heat losses. A constant load of about 4 MPa will be applied to the top and sides of the large block. The sides will be sealed with moisture and thermal barriers. Temperature, moisture content, pore pressure, chemical composition, stress, displacement, electrical resistivity, acoustic emissions, and acoustic velocities will be measured throughout the block during the heating and cool-down phases. The results from the experiments on small blocks and the tests on the large block will provide a better understanding of some concepts of the coupled TMHC processes. The progress of the project is presented in this paper

Who needs a small reactor?

International Nuclear Information System (INIS)

Wood, Janet.

1991-01-01

The opportunities and problems facing small reactors were debated at the Delhi seminar. It was established that these were markets where small reactors, producing heat as well as electricity, might be of use. Small combined heat and power reactors would be more useful in district heating than would large reactors, as their optimum heat production is in line with current district heating schemes. Most process heat requirements are below 900 o C and so may be provided by small nuclear plants. Several areas in electricity supply where small and medium sized reactors could find a market were also identified. Despite good reasons for favouring nuclear plants in these markets, such as no production of carbon dioxide, no need to use expensive oil or other scarce fossil fuels and flexibility, these are, however, disincentives to potential buyers. While serial production would decrease plant costs, the lead plants would bear heavy financial risks. Currently too many options in plant design make it difficult to present the advantages of small reactor technology. Siting reactors near centres of population would be problematical. The disposal of spent fuel and radioactive wastes would create problems in developing or non-nuclear countries. Over and above all these problems, however, was that of public acceptance. Some ways of overcoming these disincentives were discussed. (author)

The Apatity nuclear heating plant project: modern technical and economic issues of nuclear heat application in Russia

International Nuclear Information System (INIS)

Adamov, E.O.; Romenkov, A.A.

1998-01-01

Traditionally Russia is a country with advanced structure of centralized heat supply. Many thermal plants and heating networks need technical upgrading to improve their technical and economic efficiency. Fossil fuelled heating capacities have a negative influence on ecology, which can be seen especially in the northern regions of Russia. Furthermore, fossil fuel prices are rising in Russia. The above factors tend to intensify the need for alternative heat sources being capable of solving the problem. Nuclear heat sources may be the alternative. In this paper, the main features of a proposed NHP in the Murmansk region are summarized. (author)

Improvements of reforming performance of a nuclear heated steam reforming process

International Nuclear Information System (INIS)

Hada, Kazuhiko

1996-10-01

Performance of an energy production process by utilizing high temperature nuclear process heat was not competitive to that by utilizing non-nuclear process heat, especially fossil-fired process heat due to its less favorable chemical reaction conditions. Less favorable conditions are because a temperature of the nuclear generated heat is around 950degC and the heat transferring fluid is the helium gas pressurized at around 4 MPa. Improvements of reforming performance of nuclear heated steam reforming process were proposed in the present report. The steam reforming process, one of hydrogen production processes, has the possibility to be industrialized as a nuclear heated process as early as expected, and technical solutions to resolve issues for coupling an HTGR with the steam reforming system are applicable to other nuclear-heated hydrogen production systems. The improvements are as follows: As for the steam reformer, (1) increase in heat input to process gas by applying a bayonet type of reformer tubes and so on, (2) increase in reforming temperature by enhancing heat transfer rate by the use of combined promoters of orifice baffles, cylindrical thermal radiation pipes and other proposal, and (3) increase in conversion rate of methane to hydrogen by optimizing chemical compositions of feed process gas. Regarding system arrangement, a steam generator and superheater are set in the helium loop as downstream coolers of the steam reformer, so as to effectively utilize the residual nuclear heat for generating feed steam. The improvements are estimated to achieve the hydrogen production rate of approximately 3800 STP-m 3 /h for the heat source of 10 MW and therefore will provide the potential competitiveness to a fossil-fired steam reforming process. Those improvements also provide the compactness of reformer tubes, giving the applicability of seamless tubes. (J.P.N.)

Economic feasibility of heat supply from nuclear power plants in the United States

International Nuclear Information System (INIS)

Roe, K.K.; Oliker, I.

1987-01-01

Nuclear energy is regarded as competitive for urban district heating applications. Hot water heat transoport systems of up to 50 miles are feasible for heat loads over 1500 MWt, and heat load density of over 130 MWt/mi 2 is most suitable for nuclear applications. An incremental approach and a nuclear plant design provision for future heat extraction are recommended. Nuclear district heating technology status is discussed, particularly turbine design. Results of a study for retrofitting a major existing nuclear power plant to cogeneration operation are presented. The study indicates that for transmission distances up to 20 miles it is economical to generate and transport between 600 and 1200 MWt of district heat (author)

A Critical Heat Generation for Safe Nuclear Fuels after a LOCA

Directory of Open Access Journals (Sweden)

Jae-Yong Kim

2014-01-01

Full Text Available This study applies a thermo-elasto-plastic-creep finite element procedure to the analysis of an accidental behavior of nuclear fuel as well as normal behavior. The result will be used as basic data for the robust design of nuclear power plant and fuels. We extended the range of mechanical strain from small or medium to large adopting the Hencky logarithmic strain measure in addition to the Green-Lagrange strain and Almansi strain measures, for the possible large strain situation in accidental environments. We found that there is a critical heat generation after LOCA without ECCS (event category 5, under which the cladding of fuel sustains the internal pressure and temperature for the time being for the rescue of the power plant. With the heat generation above the critical value caused by malfunctioning of the control rods, the stiffness of cladding becomes zero due to the softening by high temperature. The weak position of cladding along the length continuously bulges radially to burst and to discharge radioactive substances. This kind of cases should be avoid by any means.

Development of hot water supply system for a small district heating reactor

International Nuclear Information System (INIS)

Murase, Toshihiko; Narabayashi, Tadashi; Shimazu, Yoichiro

2007-01-01

On the earth, there are many environmental problems. For example, rapid increase of world population causes the enormous consumption of fossil fuel and emission of CO 2 into the global air. Now, mankaind faced to deal with these serious problems. One solution for these problems is utilization of nuclear reactors. Currently, about 65% of thermal output of a nuclear reactor is thrown away to the sea or the atmosphere through a turbine condenser. When a hot-water pipeline from a nuclear plant will be constructed, the exhaust heat from nuclear reactor will able to be utilized. Therefore, authors began to study nuclear power plant system for district heating. This reactor is based on a PWR plant. Its thermal output is 10 MWth and its electrical output is 3.4 MW. The nuclear plant supply electricity and heat for 2000 to 3000 houses. The plant aim to supply all the energy for the adjacent pepole's life, for example, heat, electricity and hydrogen for fuel battery car. This total-energy supply system assumed to be built in Northern area such as Hokkaido in Japan. In order to develop an optimum thermal design method for the system, heat transport experiments and thermal-hydraulic calculations were carried out. Using a metal pipe covered with foam-polyurethane thermal insulator, feed-water temperature and return-water temperature was measured to evaluate heat loss. As the result, the heat loss from the hot-water temperature was very little. The thermal-hydraulic calculation method was verified and applied to actual pipeline size calculation. The result of heat loss calculation will be 0.2degC/5 km. considering these results, the best pipe specification was obtained. (author)

District heating by the Bohunice nuclear power plant

International Nuclear Information System (INIS)

Metke, E.; Skvarka, P.

1984-01-01

Technical and economical aspects of district heating by the electricity generating nuclear plants in Czechoslovakia are discussed. As a first stage of the project, 240 MW thermal power will be supplied using bleeding lines steam from the B-2 nuclear power plant at Jaslovske Bohunice to heat up water at a central station to 130 grad C. The maximal thermal power that can be produced for district heating by WWER type reactors with regular condensation turbines is estimated to be: 465 MW for a WWER-440 reactor with two 220 MWe turbines and 950 MW for a WWER-1000 reactor with a Skoda made 1000 MWe turbine using a three-stage scheme to heat up water from 60 grad C to 150 grad C. The use of satelite heating turbines connected to the steam collector is expected to improve the efficiency. District heating needs will de taken into account for siting of the new power plants

The kinetics of removal of heat-induced excess nuclear protein

International Nuclear Information System (INIS)

Roti, J.L.R.; Uygur, N.; Higashikubo, R.

1984-01-01

To investigate the role of protein content, temperature and heating time in the removal of heat-induced excess protein associated with the isolated nucleus, the kinetics of protein removal was monitored for 6 to 8 hours following exposure to 7 hyperthermic protocols. Four of these (47 0 C-7.5 min., 46 0 C-15 min., 45 0 C-30 min., and 44 0 C-60 min.) resulted in a nuclear protein content approximately twice that of nuclei from unheated cells (2.05 +- .14) following heat exposure. Three protocols (45 0 C-15 min., 44 0 C-30 min. and 43 0 C-60 min.) resulted in a nuclear protein content approximately 1.6 times normal (1.63 +- .12). If nuclear protein content were the only determinant in the recovery rate, then the same half time for nuclear protein removal would be expected within each group of protocols. Rate constants for nuclear protein removal were obtained by regression analysis. The half-time for nuclear protein removal increased with decreasing temperature and increasing heating time for the same nuclear protein content. This result suggests that the heating time and temperature are more of a determinant in the removal kinetics than protein content alone. Extended kinetics of recovery (to 36 hours) showed incomplete recovery and a secondary increase in protein associated with the isolated nucleus. These results were due to cell-cycle rearrangement (G/sub 2/ block) and unbalanced growth

Gasification of coal making use of nuclear processing heat

International Nuclear Information System (INIS)

Schilling, H.D.; Bonn, B.; Krauss, U.

1981-01-01

In the chapter 'Gasification of coal making use of nuclear processing heat', the steam gasification of brown coal and bituminous coal, the hydrogenating gasification of brown coal including nuclear process heat either by steam cracking methane in the steam reformer or by preheating the gasifying agent, as well as the hydrogenating gasification of bituminous coal are described. (HS) [de

Simulation of nuclear fuel rods by using process computer-controlled power for indirect electrically heated rods

International Nuclear Information System (INIS)

Malang, S.

1975-11-01

An investigation was carried out to determine how the simulation of nuclear fuel rods with indirect electrically heated rods could be improved by use of a computer to control the electrical power during a loss-of-coolant accident (LOCA). To aid in the experiment, a new version of the HETRAP code was developed which simulates a LOCA with heater rod power controlled by a computer that adjusts rod power during a blowdown to minimize the difference in heat flux of the fuel and heater rods. Results show that without computer control of heater rod power, only the part of a blowdown up to the time when the heat transfer mode changes from nucleate boiling to transition or film boiling can be simulated well and then only for short times. With computer control, the surface heat flux and temperature of an electrically heated rod can be made nearly identical to that of a reactor fuel rod with the same cooling conditions during much of the LOCA. A small process control computer can be used to achieve close simulation of a nuclear fuel rod with an indirect electrically heated rod

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

Energy Technology Data Exchange (ETDEWEB)

Ehud Greenspan

2008-09-30

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).

Nuclear heat generating plants - technical concepts and market potentials. Chapter 8

International Nuclear Information System (INIS)

Thoene, E.

1988-01-01

To determine the advantages and disadvantages of different heat generating systems, a comparison is made between nuclear heat generating plants and competing heat generating systems. Nuclear heat generating plant concepts in practice have to compete with a wide range of existing and new fossil heat generating technologies of the most different capacities, ranging from combined heat and power generation to individual heating in one-family houses. Heat generation costs are calculated by means of a dynamic annuity method from an economic point of view. The development of real prices of fossil energy sources is based on two scenarios characterized as follows: scenario I - insignificant price increase by the year 2000, then stagnant; scenario II - moderate price increase by the year 2010, then stagnant. As a result of that systems comparison it can be stated that the considered nuclear heat generating plants may be an interesting competitive heat generation option, provided the assumptions on which the study is based can be implemented. This applies especially to investment costs. At the same time those plants contribute to a diversification of energy source options on the heat market. Their use leads to a reduction of fossil fuel imports, increasing at the same time short- and long-term supply guarantees. If nuclear heat generating plants substitute fossil heat generating plants, or render the construction of new ones superfluous, they contribute to avoiding chemical air pollutants. (orig./UA) [de

Nuclear steam turbines for power production in combination with heating

International Nuclear Information System (INIS)

Frilund, B.; Knudsen, K.

1977-01-01

The general operating conditions for nuclear steam turbines in district heating system are briefly outlined. The turbine plant can consist of essentially the same types of machines as in conventional district heating systems. Some possible arrangements of back-pressure turbines, back-pressure turbines with condensing tails, or condensing turbines with heat extraction are considered for nuclear power and heat stations. Principles of control for hot water temperature and electrical output are described. Optimization of the plant, considering parallel variations during the year between heat load, cooling water temperature, and required outgoing temperature is discussed. (U.K.)

Five MW Nuclear Heating Reactor

International Nuclear Information System (INIS)

Zhang Dafang; Dong Duo; Su Qingshan

1997-01-01

The 5 MW Nuclear Heating Reactor (NHR-5) developed and designed by the Institute of Nuclear Energy Technology (INET) has been operated for four winter seasons since 1989. During the time of commissioning and operation a number of experiments including self-stability, self-regulation, and simulation of ATWS etc. were carried out. Some operating experiences such as water chemistry, radiation protection and environmental impacts and so on were also obtained at the same time. All of these results demonstrate the design of the NHR-5 is successful. (author). 9 refs, 11 figs, 5 tabs

Five MW Nuclear Heating Reactor

Energy Technology Data Exchange (ETDEWEB)

Dafang, Zhang; Duo, Dong; Qingshan, Su [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

1997-09-01

The 5 MW Nuclear Heating Reactor (NHR-5) developed and designed by the Institute of Nuclear Energy Technology (INET) has been operated for four winter seasons since 1989. During the time of commissioning and operation a number of experiments including self-stability, self-regulation, and simulation of ATWS etc. were carried out. Some operating experiences such as water chemistry, radiation protection and environmental impacts and so on were also obtained at the same time. All of these results demonstrate the design of the NHR-5 is successful. (author). 9 refs, 11 figs, 5 tabs.

Multipurpose nuclear process heat for energy supply in Brazil

International Nuclear Information System (INIS)

Hansen, U.; Inden, P.; Oesterwind, D.; Hukai, R.Y.; Pessine, R.T.; Pieroni, R.R.; Visoni, E.

1978-11-01

The industrialized nations require 75% of the energy as heat and it is likely that developing countries in the course of industrialization will show a comparable energy consumption structure. The High Temperature Reactor (HTR) allows the utilization of nuclear energy at high temperatures as process heat. In the Federal Republic of Germany (FRG) the development in the relevant technical areas is well advanced and warrants investigation as a matter for transfer to Brazil. In Brazil nuclear process heat finds possible applications in steel making, shale oil extraction, petroleum refining, and in the more distant future coal gasification with distribution networks. Based on growth forecasts for these industries a theoretical potential market of 38-53 GW (th) can be identified. At present nuclear process heat is marginally more expensive than conventional fossil technologies but the anticipated development is expected to add an economic incentive to the emerging necessity of providing a sound energy base in the developing countries. (author)

Atmospheric considerations regarding the impact of heat dissipation from a nuclear energy center

International Nuclear Information System (INIS)

Rotty, R.M.; Bauman, H.; Bennett, L.L.

1976-05-01

Potential changes in climate resulting from a large nuclear energy center are discussed. On a global scale, no noticeable changes are likely, but on both a regional and a local scale, changes can be expected. Depending on the cooling system employed, the amount of fog may increase, the amount and distribution of precipitation will change, and the frequency or location of severe storms may change. Very large heat releases over small surface areas can result in greater atmospheric instability; a large number of closely spaced natural-draft cooling towers have this disadvantage. On the other hand, employment of natural-draft towers makes an increase in the occurrence of ground fog unlikely. The analysis suggests that the cooling towers for a large nuclear energy center should be located in clusters of four with at least 2.5-mile spacing between the clusters. This is equivalent to the requirement of one acre of land surface per each two megawatts of heat being rejected

Nuclear-enhanced geothermal heat recovery

International Nuclear Information System (INIS)

Clark, W.H. II

1995-01-01

This report proposes the testing of an abandoned drill well for the disposal of spent nuclear fuel rods. The well need not be in a geothermal field, since the downhole assembly takes advantage of only the natural thermal gradient. The water in the immediate vicinity of the fuel will be chemically treated for corrosion resistance. Above this will be a long column of viscous fluid insoluble in water, to act as a fluid barrier. The remainder of the well bore, up to the surface, will be the working fluid for the power turbine at the surface. There will be a low-pressure region in the immediate vicinity of the fuel, encouraging the flashing of steam. Due to the low level of heat emitted by the fuel rods, the radioactive material will be surrounded by a secondary casing that will reduce the water it contacts directly, thus causing it to heat up quickly and to maximize the steam-generating process, and the formation of air nuclides. These will percolate upward through the viscous column where steadily decreasing pressure causes expansion. The nuclear fuel's thermal energy will have been transferred through the high radioactive zone as pressure, then it will flash to steam and heat the water in the top of the wellbore. The drill well, a minimum of 10,000 ft. in depth, will naturally heat any circulating fluid. The fuel is not used as a thermal source, but only to produce a few spontaneous bubbles, sufficient to increase the fluid pressure by expansion as it rises in the wellbore. The additional thermal energy from the nuclear source will superheat the water for use in the power-generation apparatus at the surface. This equipment, operating on very-low radioactive fluid, will be protected by a secondary containment. The typical drill well is ideally suited for the insertion of spent fuel rods, which are smaller than downhole tools and instrumentation regularly installed in production wells

Oil shales and the nuclear process heat

International Nuclear Information System (INIS)

Scarpinella, C.A.

1974-01-01

Two of the primary energy sources most dited as alternatives to the traditional fossil fuels are oil shales and nuclear energy. Several proposed processes for the extraction and utilization of oil and gas from shale are given. Possible efficient ways in which nuclear heat may be used in these processes are discussed [pt

Small reactors and the 'second nuclear era'

International Nuclear Information System (INIS)

Egan, J.R.

1984-01-01

Predictions of the nuclear industry's demise are premature and distort both history and politics. The industry is reemerging in a form commensurate with the priorities of those people and nations controlling the global forces of production. The current lull in plant orders is due primarily to the world recession and to factors related specifically to reactor size. Traditional economies of scale for nuclear plants have been greatly exaggerated. Reactor vendors and governments in Great Britain, France, West Germany, Japan, the United States, Sweden, Canada, and the Soviet Union are developing small reactors for both domestic applications and export to the Third World. The prefabricated, factory-assembled plants under 500 MWe may alleviate many of the existing socioeconomic constraints on nuclear manufacturing, construction, and operation. In the industrialized world, small reactors could furnish a qualitatively new energy option for utilities. But developing nations hold the largest potential market for small reactors due to the modest size of their electrical systems. These units could double or triple the market potential for nuclear power in this century. Small reactors will both qualitatively and quantitatively change the nature of nuclear technology transfers, offering unique advantages and problems vis-a-vis conventional arrangements. (author)

Perspectives of heat transfer enhancement in nuclear reactors toward nanofluids applications

International Nuclear Information System (INIS)

Rocha, Marcelo S.; Cabral, Eduardo L.L.; Sabundjian, Gaiane

2013-01-01

Nanofluids are colloidal suspensions of nanoparticles in a base fluid with interesting physical properties and large potential for heat transfer enhancement in thermal systems among other applications. There are an increasing number of nanofluids investigations concerning many aspects of synthesis and fabrication technologies, physical properties, and special applications. Results demonstrate that physical properties like high thermal conductivities and high critical heat flux (CHF) of some nanofluids classifies them as potential working fluids for high heat flux transportation in special systems, including thermal management of microelectronic devices (MEMS) and nuclear reactors. Understanding the importance of such investigations for the knowledge development of nuclear engineering a new research is being conducted at the Nuclear Engineering Center (CEN) of the Nuclear and Energy Research Institute (IPEN/CNEN-SP) to analyze the application potentiality of some nanofluids in nuclear systems for heat transfer enhancement under ionizing radiation influence. In this work a revision of theoretical and experimental studies of nanofluids is performed and its potentiality for using in future generations of nuclear reactors is highlighted showing the status of the research at present. (author)

Method and apparatus for nuclear heating of oil-bearing formations

International Nuclear Information System (INIS)

Alspaw, D.I.

1979-01-01

A method and apparatus are provided for using heat generated by absorption of radiation from nuclear waste materials to reduce the viscosity of petroleum products contained within a subsurface earth formation. The nuclear waste material is positioned in a salt water formation underlying the subsurface earth formation so that the radiation emitted by the material heats the salt water formation. conduction and convection transfer the heat to the subsurface earth formation, raising the temperature and thereby reducing the viscosity of the petroleum products. To prevent radioactive contamination within the salt water formation, the nuclear waste material may be encapsulated in a material selected to absorb alpha and beta radiation

Heat transfer characteristics and operation limit of pressurized hybrid heat pipe for small modular reactors

International Nuclear Information System (INIS)

Kim, Kyung Mo; Bang, In Cheol

2017-01-01

Highlights: • Thermal performances and operation limits of hybrid heat pipe were experimentally studied. • Models for predicting the operation limit of the hybrid heat pipe was developed. • Non-condensable gas affected heat transfer characteristics of the hybrid heat pipe. - Abstract: In this paper, a hybrid heat pipe is proposed for use in advanced nuclear power plants as a passive heat transfer device. The hybrid heat pipe combines the functions of a heat pipe and a control rod to simultaneously remove the decay heat generated from the core and shutdown the reactor under accident conditions. Thus, the hybrid heat pipe contains a neutron absorber in the evaporator section, which corresponds to the core of the reactor pressure vessel. The presence of the neutron absorber material leads to differences in the heated diameter and hydraulic diameter of the heat pipe. The cross-sectional areas of the vapor paths through the evaporator, adiabatic, and condenser sections are also different. The hybrid heat pipe must operate in a high-temperature, high-pressure environment to remove the decay heat. In other words, the operating pressure must be higher than those of the commercially available thermosyphons. Hence, the thermal performances, including operation limit of the hybrid heat pipe, were experimentally studied in the operating pressure range of 0.2–20 bar. The operating pressure of the hybrid heat pipe was controlled by charging the non-condensable gas which is unused method to achieve the high saturation pressure in conventional thermosyphons. The effect of operating pressure on evaporation heat transfer was negligible, while condensation heat transfer was affected by the amount of non-condensable gas in the test section. The operation limit of the hybrid heat pipe increased with the operating pressure. Maximum heat removal capacity of the hybrid heat pipe was up to 6 kW which is meaningful value as a passive decay heat removal device in the nuclear power

Thermal behavior simulation of a nuclear fuel rod through an eletrically heated rod

International Nuclear Information System (INIS)

Lima, R. de C.F. de.

1984-01-01

In thermalhydraulic loops the nuclear industry often uses electrically heated rods to simulate power transients, which occur in nuclear fuel rods. The development and design of a electrically heated rod, by supplying the dimensions and materials which should be used in order to yeld the same temperature and heat flux at the surfaces of the nuclear rod and the electrically heated rod are presented. To a given nuclear transient this equality was obtained by fitting the linear power through the lumped parameters technique. (Author) [pt

Acceleration of calculation of nuclear heating distributions in ITER toroidal field coils using hybrid Monte Carlo/deterministic techniques

International Nuclear Information System (INIS)

Ibrahim, Ahmad M.; Polunovskiy, Eduard; Loughlin, Michael J.; Grove, Robert E.; Sawan, Mohamed E.

2016-01-01

Highlights: • Assess the detailed distribution of the nuclear heating among the components of the ITER toroidal field coils. • Utilize the FW-CADIS method to dramatically accelerate the calculation of detailed nuclear analysis. • Compare the efficiency and reliability of the FW-CADIS method and the MCNP weight window generator. - Abstract: Because the superconductivity of the ITER toroidal field coils (TFC) must be protected against local overheating, detailed spatial distribution of the TFC nuclear heating is needed to assess the acceptability of the designs of the blanket, vacuum vessel (VV), and VV thermal shield. Accurate Monte Carlo calculations of the distributions of the TFC nuclear heating are challenged by the small volumes of the tally segmentations and by the thick layers of shielding provided by the blanket and VV. To speed up the MCNP calculation of the nuclear heating distribution in different segments of the coil casing, ground insulation, and winding packs of the ITER TFC, the ITER Organization (IO) used the MCNP weight window generator (WWG). The maximum relative uncertainty of the tallies in this calculation was 82.7%. In this work, this MCNP calculation was repeated using variance reduction parameters generated by the Oak Ridge National Laboratory AutomateD VAriaNce reducTion Generator (ADVANTG) code and both MCNP calculations were compared in terms of computational efficiency and reliability. Even though the ADVANTG MCNP calculation used less than one-sixth of the computational resources of the IO calculation, the relative uncertainties of all the tallies in the ADVANTG MCNP calculation were less than 6.1%. The nuclear heating results of the two calculations were significantly different by factors between 1.5 and 2.3 in some of the segments of the furthest winding pack turn from the plasma neutron source. Even though the nuclear heating in this turn may not affect the ITER design because it is much smaller than the nuclear heating in the

Acceleration of calculation of nuclear heating distributions in ITER toroidal field coils using hybrid Monte Carlo/deterministic techniques

Energy Technology Data Exchange (ETDEWEB)

Ibrahim, Ahmad M., E-mail: ibrahimam@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Polunovskiy, Eduard; Loughlin, Michael J. [ITER Organization, Route de Vinon Sur Verdon, 13067 St. Paul Lez Durance (France); Grove, Robert E. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Sawan, Mohamed E. [University of Wisconsin-Madison, 1500 Engineering Dr., Madison, WI 53706 (United States)

2016-11-01

Highlights: • Assess the detailed distribution of the nuclear heating among the components of the ITER toroidal field coils. • Utilize the FW-CADIS method to dramatically accelerate the calculation of detailed nuclear analysis. • Compare the efficiency and reliability of the FW-CADIS method and the MCNP weight window generator. - Abstract: Because the superconductivity of the ITER toroidal field coils (TFC) must be protected against local overheating, detailed spatial distribution of the TFC nuclear heating is needed to assess the acceptability of the designs of the blanket, vacuum vessel (VV), and VV thermal shield. Accurate Monte Carlo calculations of the distributions of the TFC nuclear heating are challenged by the small volumes of the tally segmentations and by the thick layers of shielding provided by the blanket and VV. To speed up the MCNP calculation of the nuclear heating distribution in different segments of the coil casing, ground insulation, and winding packs of the ITER TFC, the ITER Organization (IO) used the MCNP weight window generator (WWG). The maximum relative uncertainty of the tallies in this calculation was 82.7%. In this work, this MCNP calculation was repeated using variance reduction parameters generated by the Oak Ridge National Laboratory AutomateD VAriaNce reducTion Generator (ADVANTG) code and both MCNP calculations were compared in terms of computational efficiency and reliability. Even though the ADVANTG MCNP calculation used less than one-sixth of the computational resources of the IO calculation, the relative uncertainties of all the tallies in the ADVANTG MCNP calculation were less than 6.1%. The nuclear heating results of the two calculations were significantly different by factors between 1.5 and 2.3 in some of the segments of the furthest winding pack turn from the plasma neutron source. Even though the nuclear heating in this turn may not affect the ITER design because it is much smaller than the nuclear heating in the

Target study of heat supply from Northern Moravia nuclear power plant

International Nuclear Information System (INIS)

Pospisil, V.

The construction is envisaged in Northern Moravia of a nuclear power plant near Blahutovice in the Novy Jicin district. Heat produced by the nuclear power plant will only be used for district heating; process heat will be supplied from local steam sources. An example is discussed of the Prerov locality which currently is supplied from the Prerov heating and power plant (230 MW), a heating plant (36 MW) and from local sources (15 NW). The study estimates that a thermal power of 430 MW will be required at a time of the start of heat supplies from the nuclear power plant. All heat supply pipelines will be designed as a two-tube system divided into sections with section pipe fittings. The number and location of pipe fittings will be selected depending on the terrain configuration. Water of the maximum outlet temperature of 150 degC will be used as a coolant. The control of the system for Northern Moravia is briefly described. (J.P.)

After-heat removing device in nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Mizuno, K [Nippon Atomic Industry Group Co. Ltd., Tokyo

1977-01-14

Purpose: To prevent water hammer in a BWR type reactor or the like by moving water in pipe lines having stagnant portions in an after-heat removing device. Constitution: To a reactor container, is provided a recycling pump which constitutes a closed loop type recycling system in a nuclear power plant together with a pressure vessel and pipe lines. A pump and a heat exchanger are provided outside of the reactor container and they are connected to up- and down-streams of the recycling pump to form an after-heat removing device in the plant. Upon shutdown of the nuclear power plant, since water in the stagnant portion flows to the intake port of the recycling pump and water from the reactor is spontaneously supplemented thereafter to the stagnant portion, neither pressurized water nor heated steam is generated and thus water hammer is prevented.

Method of calculating heat transfer in furnaces of small power

Directory of Open Access Journals (Sweden)

Khavanov Pavel

2016-01-01

Full Text Available This publication presents the experiences and results of generalization criterion equation of importance in the analysis of the processes of heat transfer and thermal calculations of low-power heat generators cooled combustion chambers. With generalizing depending estimated contribution of radiation and convective heat transfer component in the complex for the combustion chambers of small capacity boilers. Determined qualitative and quantitative dependence of the integrated radiative-convective heat transfer from the main factors working combustion chambers of small volume.

Dedicated low temperature nuclear district heating plants: Rationale and prospects

International Nuclear Information System (INIS)

Goetzmann, C.A.

1997-01-01

Space heating accounts for a substantial fraction of the end-energy consumption in a large number of industrialized countries. Accordingly, efforts have been under way since many years to utilize nuclear energy as a source for district heating. The paper describes the key technical and institutional issues affecting the implementation of such technology. It is argued that the basic case for nuclear district heating is sound but that its introduction merits and drawbacks strongly depend on local circumstances. (author). 4 figs, 1 tab

Dedicated low temperature nuclear district heating plants: Rationale and prospects

Energy Technology Data Exchange (ETDEWEB)

Goetzmann, C A [Division of Nuclear Power, International Atomic Energy Agency, Vienna (Austria)

1997-09-01

Space heating accounts for a substantial fraction of the end-energy consumption in a large number of industrialized countries. Accordingly, efforts have been under way since many years to utilize nuclear energy as a source for district heating. The paper describes the key technical and institutional issues affecting the implementation of such technology. It is argued that the basic case for nuclear district heating is sound but that its introduction merits and drawbacks strongly depend on local circumstances. (author). 4 figs, 1 tab.

78 FR 55117 - Ultimate Heat Sink for Nuclear Power Plants; Draft Regulatory Guide

Science.gov (United States)

2013-09-09

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0203] Ultimate Heat Sink for Nuclear Power Plants; Draft... (DG), DG-1275, ``Ultimate Heat Sink for Nuclear Power Plants.'' This regulatory guide (RG) describes methods and procedures acceptable to the NRC staff that nuclear power plant facility licensees and...

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.

The Swiss heating reactor (SHR) for district heating of small communities

International Nuclear Information System (INIS)

Burgsmueller, P.; Jacobi, A.Jr.; Jaeger, J.F.; Klaentschi, M.J.; Seifritz, W.; Vuillemier, F.; Wegmann, F.

1987-01-01

With fossil fuel running out in a foreseeable future, it is essential to develop substitution strategies. Some 40-50 % of the heat demand in industrial countries is below 120 degrees C, for space heating and warm water production, causing a corresponding fraction of air pollution by SO 2 and to a lesser extent NO x if fossil fuels are used. Yet, contemporary LWR technology makes it feasible to supply a district heating network without basically new reactor development. Units in the power range 10-50 MW are most suitable for Switzerland, both in respect of network size and of the democratic decision making structure. A small BWR for heating purpose is being developed by parts of the Swiss Industry and the Swiss Federal Institute for Reactor Research (EIR). The economic target of 100-120 SFr/MWh heat at the consumer's seems achievable. (author)

Process for extracting residual heat and device for the ultimate absorption of heat for nuclear reactors

International Nuclear Information System (INIS)

Bernard, Lawrence Jr.

1980-01-01

This invention concerns a 'heat sink' or device for the ultimate absorption of heat for electric power stations using the most widespread thermal neutron nuclear reactors, namely 'light water' reactors such as boiling or pressurized water reactors. The residual heat given off by these reactors can be safely extracted with this method by using dry cooling. However, the invention does not concern the problems arising from the cooling of the steam used for actuating the steam turbine nor the cooling of the steam exhausted by the turbine or coming from it, but it does concern the 'safety' part of the nuclear power station in which the residual heat discharged in the reactor is controlled and dissipated [fr

Advances in Nuclear Power Process Heat Applications

International Nuclear Information System (INIS)

2012-05-01

Following an IAEA coordinated research project, this publication compiles the findings of research and development activities related to practical nuclear process heat applications. An overview of current progress on high temperature gas cooled reactors coupling schemes for different process heat applications, such as hydrogen production and desalination is included. The associated safety aspects are also highlighted. The summary report documents the results and conclusions of the project.

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

SLOWPOKE: heating reactors in the urban environment

International Nuclear Information System (INIS)

Hilborn, J.W.; Lynch, G.F.

1988-06-01

Since global energy requirements are expected to double over the next 40 years, nuclear heating could become as important as nuclear electricity generation. To fill that need, AECL has designed a 10 MW nuclear heating plant for large buildings. Producing hot water at temperatures below 100 degrees Celsius, it incorporates a small pool-type reactor based on the successful SLOWPOKE Research Reactor. A 2 MW prototype is now being tested at the Whiteshell Nuclear Research Establishment in Manitoba, and the design of a 10 MW commercial unit is well advanced. With capital costs in the range $5 million to $7 million, unit energy costs could be as low as $0.02 per kWh, for a unit operating at 50% load factor over a 25-year period. By keeping the reactor power low and the water temperature below 100 degrees Celsius, 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

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

Hydrogen and oxygen production with nuclear heat

International Nuclear Information System (INIS)

Barnert, H.

1979-09-01

After some remarks on the necessity of producing secondary energy sources for the heat market, the thermodynamic fundamentals of the processes for producing hydrogen and oxygen from water on the basis of nuclear thermal energy are briefly explained. These processes are summarized as one class of the 'thermochemical cycle process' for the conversion of thermal into chemical energy. A number of thermochemical cycle processes are described. The results of the design work so far are illustrated by the example of the 'sulphuric acid hybrid process'. The nuclear heat source of the thermochemical cycle process is the high-temperature reactor. Statements concerning rentability are briefly commented upon, and the research and development efforts and expenditure required are sketched. (orig.) 891 GG/orig. 892 MB [de

Design and application for a high-temperature nuclear heat source

International Nuclear Information System (INIS)

Quade, R.N.

1980-01-01

Recent actions by OPEC have sharply increased interest in the United States in synfuels, with coal being the logical choice for the carbon source. Two coal liquefaction processes, direct and indirect, have been examined. Each can produce about 50% more output when coupled to an HTGR for process heat. The nuclear reactor designed for process heat has a power output of 842MW(t), a core outlet temperature of 950 0 C (1742 0 F), and an intermediate helium loop to separate the heat source from the process heat exchangers. Steam-methane reforming is the reference process. As part of the development of a nuclear process heat system, a computer code, Process Heat Reactor Evaluation and Design, is being developed. This code models both the reactor plant and a steam reforming plant. When complete, the program will have the capability to calculate an overall mass and heat balance, size the plant components, and estimate the plant cost for a wide variety of independent variables. (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.

Nuclear source of district heating in the north-east region of Russia

International Nuclear Information System (INIS)

Dolgov, V.V.

1998-01-01

The operation of the Bilibin Nuclear Co-generation Plant (BNCP) as a local district heating source is reviewed in this paper. Specific features of the BNCP power unit are given with special emphases on the components of the technological scheme, which are involved in the heat production and supply to the consumers. The scheme of steam extraction from the turbine, the flow diagram of steam in the turbine, as well as the three circuit heat removal system are described. The numerical characteristics of the nuclear heat supply system in various operating modes are presented. The real information characterizing current radiological conditions in the vicinity of the heat generation and distribution equipment is also presented in the paper. The BNCP technical and economical characteristics are compared with those of conventional energy sources. Both advantages and some problems revealed during the twenty-year experience of the BNCP nuclear heat utilization are generally assessed. Safety and reliability characteristics of the reactor and the heat supply system are also described. (author)

Small nuclear reactor safety design requirements for autonomous operation

International Nuclear Information System (INIS)

Kozier, K.S.; Kupca, S.

1997-01-01

Small nuclear power reactors offer compelling safety advantages in terms of the limited consequences that can arise from major accident events and the enhanced ability to use reliable, passive means to eliminate their occurrence by design. Accordingly, for some small reactor designs featuring a high degree of safety autonomy, it may be-possible to delineate a ''safety envelope'' for a given set of reactor circumstances within which safe reactor operation can be guaranteed without outside intervention for time periods of practical significance (i.e., days or weeks). The capability to operate a small reactor without the need for highly skilled technical staff permanently present, but with continuous remote monitoring, would aid the economic case for small reactors, simplify their use in remote regions and enhance safety by limiting the potential for accidents initiated by inappropriate operator action. This paper considers some of the technical design options and issues associated with the use of small power reactors in an autonomous mode for limited periods. The focus is on systems that are suitable for a variety of applications, producing steam for electricity generation, district heating, water desalination and/or marine propulsion. Near-term prospects at low power levels favour the use of pressurized, light-water-cooled reactor designs, among which those having an integral core arrangement appear to offer cost and passive-safety advantages. Small integral pressurized water reactors have been studied in many countries, including the test operation of prototype systems. (author)

The Economics of Connecting of Small Buildings to Geothermal District Heating Systems

Energy Technology Data Exchange (ETDEWEB)

Rafferty, Kevin

2003-03-01

Many of the communities co-located with geothermal resources are very small and as a result the buildings they contain tend to be small as well. Generally, small buildings (10,000 ft2) use heating systems which are not hot water based. Since geothermal district heating systems deliver hot water, the costs associated with the conversion of small building heating systems to use hot water for heating is an issue of great influence in terms of the potential development of such systems. This paper examines the typical retrofit costs associated with conversion of small buildings and the level of savings necessary to attract the interest of owners. In general, the prospects for connection of such buildings based only on energy savings is not positive.

The economics of connecting of small buildings to geothermal district heating systems

Energy Technology Data Exchange (ETDEWEB)

Rafferty, Kevin

2001-01-01

Many of the communities co-located with geothermal resources are very small and as a result the buildings they contain tend to be small as well. Generally, small buildings (10,000 ft2) use heating systems which are not hot water based. Since geothermal district heating systems deliver hot water, the costs associated with the conversion of small building heating systems to use hot water for heating is an issue of great influence in terms of the potential development of such systems. This paper examines the typical retrofit costs associated with conversion of small buildings and the level of savings necessary to attract the interest of owners. In general, the prospects for connection of such buildings based only on energy savings is not positive.

Numerical investigation of heat transfer effects in small wave rotor

International Nuclear Information System (INIS)

Deng, Shi; Okamoto, Koji; Teramoto, Susumu

2015-01-01

Although a wave rotor is expected to enhance the performance of the ultra-micro gas turbine, the device itself may be affected by downsizing. Apart from the immediate effect of viscosity on flow dynamics when downscaled, the effects of heat transfer on flow field increase at such small scales. To gain an insight into the effects of heat transfer on the internal flow dynamics, numerical investigations were carried out with adiabatic, isothermal and conjugate heat transfer boundary treatments at the wall, and the results compared and discussed in the present study. With the light shed by the discussion of adiabatic and conjugate heat transfer boundary treatments, this work presents investigations of the heat flux distributions, as well as the effects of heat transfer on the internal flow dynamics and the consequent charging and discharging processes for various sizes. When heat transfer is taken into account, states of fluid in the cell before compression process varies, shock waves in compression process are found to be weaker, and changes in the charging and discharging processes are observed. Heat transfer differences between conjugate heat transfer boundary treatment and isothermal boundary treatment are addressed through comparisons of local wall temperature and heat flux. As a result, the difference in discharging temperature of high pressure fluid is noticeable in all sizes investigated, and the rapid increase of differences between results of isothermal and conjugate heat transfer boundary treatment in small size reveals that for certain small sizes (length of cell < 23 mm) the thermal boundary treatment should be taken care of.

A five MW nuclear heating reactor

International Nuclear Information System (INIS)

Zhang Dafang; Don Duo; Su Quingshan

1997-01-01

The 5 MW Nuclear Heating Reactor (NHR-5) developed and designed by the Institute of Nuclear Energy and Technology (INET) and has been operated for four winter seasons since 1989. During the time of commissioning and operation a number of experiments including self-stability, self-regulation and simulation of ATWS etc. were carried out. Some operating experiences such as water chemistry, radiation protection, and environmental impacts and so on, were also obtained at the same time. All of these demonstrate that the design of NHR-5 is successful. (author)

The Thermos program for nuclear reactors specialized in district heating

International Nuclear Information System (INIS)

Lerouge, B.

1976-01-01

Many studies have been made in France on the use of nuclear heat for district heating. After a brief account of the problems raised by the use of thermal waste from big nuclear power stations, the quantitative and qualitative needs of heating networks are analyzed and the Thermos project described. This is a very robust reactor of the pool type, with an output of 100MW, supplying low-pressure water at 100 deg C. The advantages from the aspects of safety and economy are described, and the present state of the project and its possible developments summarized [fr

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

The encapsulated nuclear heat source reactor for proliferation-resistant nuclear energy

International Nuclear Information System (INIS)

Brown, N.W.; Hossain, Q.; Carelli, M.D.; Conway, L.; Dzodzo, M.; Greenspan, E.; Saphier, D.

2001-01-01

The encapsulated nuclear heat source (ENHS) is a modular reactor that was selected by the 1999 DOE NERI program as a candidate ''Generation-IV'' reactor concept. It is a fast neutron spectrum reactor cooled by Pb-Bi using natural circulation. It is designed for passive load following, for high level of passive safety, and for 15 years without refueling. One of the unique features of the ENHS is that the fission-generated heat is transferred from the primary coolant to the secondary coolant across the reactor vessel wall by conduction-providing for an essentially sealed module that is easy to install and replace. Because the fuel is encapsulated within a heavy steel container throughout its life it provides a unique improvement to the proliferation resistance of the nuclear fuel cycle. This paper presents the innovative technology of the ENHS. (author)

A microstructural comparison of two nuclear-grade martensitic steels using small-angle neutron scattering

Science.gov (United States)

Coppola, R.; Fiori, F.; Little, E. A.; Magnani, M.

1997-06-01

Results are presented of a small-angle neutron scattering (SANS) study on two 10-13% Cr martensitic stainless steels of interest for nuclear applications, viz. DIN 1.4914 (MANET specification, for fusion reactors) and AISI 410. The investigation has focussed principally on microstructural effects associated with the differences in chromium content between the two alloys. The size distribution functions determined from nuclear and magnetic SANS components for the two steels given identical heat treatments are in accord with an interpretation based on the presence of ˜ 1 nm size CCr aggregates in the microstructure. Much larger (˜ 10 nm) scattering inhomogeneities with different magnetic contrast are also present and tentatively identified as carbides.

Cost and primary energy efficiency of small-scale district heating systems

International Nuclear Information System (INIS)

Truong, Nguyen Le; Gustavsson, Leif

2014-01-01

Highlights: • We analyzed minimum-cost options for small-scale DHSs under different contexts. • District heat production cost increases with reduced DHS scales. • Fewer technical options are suitable for small-scale DHSs. • Systems with combined technologies are less sensitive to changes in fuel prices. - Abstract: Efficient district heat production systems (DHSs) can contribute to achieving environmental targets and energy security for countries that have demands for space and water heating. The optimal options for a DHS vary with the environmental and social-political contexts and the scale of district heat production, which further depends on the size of the community served and the local climatic conditions. In this study, we design a small-scale, minimum-cost DHS that produces approximately 100 GWh heat per year and estimate the yearly production cost and primary energy use of this system. We consider conventional technologies, such as heat-only boilers, electric heat pumps and combined heat and power (CHP) units, as well as emerging technologies, such as biomass-based organic Rankine cycle (BORC) and solar water heating (SWH). We explore how different environmental and social-political situations influence the design of a minimum-cost DHS and consider both proven and potential technologies for small-scale applications. Our calculations are based on the real heat load duration curve for a town in southern Sweden. We find that the district heat production cost increases and that the potential for cogeneration decreases with smaller district heat production systems. Although the selection of technologies for a minimum-cost DHS depends on environmental and social-political contexts, fewer technical options are suitable for small-scale systems. Emerging technologies such as CHP-BORC and SWH improve the efficiency of primary energy use for heat production, but these technologies are more costly than conventional heat-only boilers. However, systems with

National need for utilizing nuclear energy for process heat generation

International Nuclear Information System (INIS)

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

1984-01-01

Nuclear reactors are potential sources for generating process heat, and their applications for such use economically competitive. They help satisfy national needs by helping conserve and extend oil and natural gas resources, thus reducing energy imports and easing future international energy concerns. Several reactor types can be utilized for generating nuclear process heat; those considered here are light water reactors (LWRs), heavy water reactors (HWRs), gas-cooled reactors (GCRs), and liquid metal reactors (LMRs). LWRs and HWRs can generate process heat up to 280 0 C, LMRs up to 540 0 C, and GCRs up to 950 0 C. Based on the studies considered here, the estimated process heat markets and the associated energy markets which would be supplied by the various reactor types are summarized

Engineering and economic aspects of centalized heating from nuclear boilers

International Nuclear Information System (INIS)

Emel'yanov, I.Ya.; Baturov, B.B.; Korytnikov, V.P.; Koryakin, Yu.I.; Chernyaev, V.A.; Kovylyanskij, Ya.A.; Galaktionov, I.V.

1979-01-01

Some engineering and economic aspects for deployment of centralized nuclear boilers (NB) in the USSR are considered. Engineering, maintenance and economic features of NB as compared to organic-fuelled boilers and nuclear thermal power plants are discussed. Among major factors governing economic efficiency of NB underlined are oraganic fuel costs, reactor unit power, location relative to heat-consuming centres and capacity factor. It is concluded that NB can be economical for heating large consumers (more than 1500 G kal/hr). At the periphery NB can be competitive already at reactor unit power of several MWth. The development of HTGR type reactor-based nuclear-chemical boilers and lines for heat transport in a chemically bound state (e.g., CH 4 → H 2 +CO 2 +CO → CH 4 ) opens the way for a substantial breakthrow in the centralized NB efficiency

Measurement of specific heat and specific absorption rate by nuclear magnetic resonance

Energy Technology Data Exchange (ETDEWEB)

Gultekin, David H., E-mail: david.gultekin@aya.yale.edu [Department of Electrical Engineering, Yale University, New Haven, CT 06520 (United States); Department of Medical Physics, Memorial Sloan-Kettering Cancer Center, New York, NY 10065 (United States); Department of Radiology, Memorial Sloan-Kettering Cancer Center, New York, NY 10065 (United States); Institute of Imaging Science, Vanderbilt University, Nashville, TN 37232 (United States); Gore, John C. [Department of Biomedical Engineering, Vanderbilt University, Nashville, TN 37232 (United States); Department of Radiology and Radiological Sciences, Vanderbilt University, Nashville, TN 37232 (United States); Department of Molecular Physiology and Biophysics, Vanderbilt University, Nashville, TN 37232 (United States); Department of Physics and Astronomy, Vanderbilt University, Nashville, TN 37232 (United States); Institute of Imaging Science, Vanderbilt University, Nashville, TN 37232 (United States)

2010-05-20

We evaluate a nuclear magnetic resonance (NMR) method of calorimetry for the measurement of specific heat (c{sub p}) and specific absorption rate (SAR) in liquids. The feasibility of NMR calorimetry is demonstrated by experimental measurements of water, ethylene glycol and glycerol using any of three different NMR parameters (chemical shift, spin-spin relaxation rate and equilibrium nuclear magnetization). The method involves heating the sample using a continuous wave laser beam and measuring the temporal variation of the spatially averaged NMR parameter by non-invasive means. The temporal variation of the spatially averaged NMR parameter as a function of thermal power yields the ratio of the heat capacity to the respective nuclear thermal coefficient, from which the specific heat can be determined for the substance. The specific absorption rate is obtained by subjecting the liquid to heating by two types of radiation, radiofrequency (RF) and near-infrared (NIR), and by measuring the change in the nuclear spin phase shift by a gradient echo imaging sequence. These studies suggest NMR may be a useful tool for measurements of the thermal properties of liquids.

Nuclear power for district heating

International Nuclear Information System (INIS)

Lyon, R.B.; Sochaski, R.O.

1975-09-01

Current district heating trends are towards an increasing use of electricity. This report concerns the evaluation of an alternative means of energy supply - the direct use of thermal energy from CANDU nuclear stations. The energy would be transmitted via a hot fluid in a pipeline over distances of up to 40 km. Advantages of this approach include a high utilization of primary energy, with a consequent reduction in installed capacity, and load flattening due to inherent energy storage capacity and transport delays. Disadvantages include the low load factors for district heating, the high cost of the distribution systems and the necessity for large-scale operation for economic viability. This requirement for large-scale operation from the beginning could cause difficulty in the implementation of the first system. Various approaches have been analysed and costed for a specific application - the supply of energy to a district heating load centre in Toronto from the location of the Pickering reactor station about 40 km away. (author)

Gasification of coal using nuclear process heat. Chapter D

International Nuclear Information System (INIS)

Schilling, H.-D.; Bonn, B.; Krauss, U.

1979-01-01

In the light of the high price of coal and the enormous advances made recently in nuclear engineering, the possibility of using heat from high-temperature nuclear reactors for gasification processes was discussed as early as the 1960s. The advantages of this technology are summarized. A joint programme of development work is described, in which the Nuclear Research Centre at Juelich is aiming to develop a high-temperature reactor which will supply process heat at as high a temperature as possible, while other organizations are working on the hydrogasification of lignites and hard coals, and steam gasification. Experiments are at present being carried out on a semi-technical scale, and no operational data for large-scale plants are available as yet. (author)

Heating of water by nuclear power stations

International Nuclear Information System (INIS)

1974-01-01

The aim of this note is to examine: the thermal conditions of the Rhone in its present state; heating caused by the building of nuclear power stations; the main hydrobiological and ecological characteristics of the Rhone [fr

Development of small and medium reactors for power and heat production

International Nuclear Information System (INIS)

Becka, J.

1978-01-01

Data are given on the current state of development of small and medium-power reactors designed mainly for electric power production in small power grids, for heat production for small- and medium-power desalination plants with possible electric power generation, for process steam production and heat development for district heating systems, again combined with electric power generation, and for propelling big and fast passenger ships. A diagram is shown of the primary system of an integrated PWR derived from the Otto Hahn reactor. The family is listed of the standard sizes of the integral INTERATOM company pressurized water reactors. Also listed are the specifications and design of CAS 2CG and AS 3G type reactors used mainly for long-distance heating systems. (J.B.)

HTGR nuclear heat source component design and experience

International Nuclear Information System (INIS)

Peinado, C.O.; Wunderlich, R.G.; Simon, W.A.

1982-05-01

The high-temperature gas-cooled reactor (HTGR) nuclear heat source components have been under design and development since the mid-1950's. Two power plants have been designed, constructed, and operated: the Peach Bottom Atomic Power Station and the Fort St. Vrain Nuclear Generating Station. Recently, development has focused on the primary system components for a 2240-MW(t) steam cycle HTGR capable of generating about 900 MW(e) electric power or alternately producing high-grade steam and cogenerating electric power. These components include the steam generators, core auxiliary heat exchangers, primary and auxiliary circulators, reactor internals, and thermal barrier system. A discussion of the design and operating experience of these components is included

Comparative study for endenergy supply with nuclear district heating and with nuclear long distance energy

International Nuclear Information System (INIS)

Dietrich, G.

1975-07-01

The future energy supply of the Federal Republic of Germany will be orientated to secure energy carriers. Moreover economical energy consumption and environmental protection will be a force for an increased application of district heating and nuclear long distance energy. The technics of generation, transport and distribution of the two energy carriers will be discussed, besides a short review of application areas and potentials. The cost comparisons by models show that there are special advantages for both systems. Nevertheless the conclusions from the study can be to favour nuclear long distance energy because of its wide application range in the whole heat market. But there is also the competition with combined heat and power generation on fossil basis, as practised in many industrial companies. As a result of a regional analysis of the area Aachen-Moenchengladbach-Koeln, the cost advantages of the nuclear long distance energy as a parameter of current prices are confirmed. Nuclear long distance energy, in combination with the high temperature reactor and a developed technic of catalysts up to temperatures of 900 K, is an energy source which will be independant of regional necessities, secure, non pollutant and economic. (orig.) [de

Laser pulse heating of nuclear fuels for simulation of reactor power

Indian Academy of Sciences (India)

Laser applications; nuclear fuel elements; nuclear safety. ... accident (LOCA) and reactivity initiated accident (RIA), a laser pulse heating system is under ... As a prelude to work on irradiated nuclear fuel specimens, pilot studies on unirradiated ...

Design of an automatic control system of a district heating nuclear plant

International Nuclear Information System (INIS)

Zebiri, Abderrahim.

1980-06-01

This paper presents the synthesis of the control system of a nuclear/oil fuelled district heating plant. Operating criteria take into account the economical background of the problem. Nuclear reactor control loops were specially conceived, due to the specific perturbations to which is submitted a district heating plant [fr

Subcooled flow boiling heat transfer from microporous surfaces in a small channel

International Nuclear Information System (INIS)

Yan, Sun; Li, Zhang; Hong, Xu; Xiaocheng, Zhong

2011-01-01

The continuously increasing requirement for high heat transfer rate in a compact space can be met by combining the small channel/microchannel and heat transfer enhancement methods during fluid subcooled flow boiling. In this paper, the sintered microporous coating, as an efficient means of enhancing nucleate boiling, was applied to a horizontal, rectangular small channel. Water flow boiling heat transfer characteristics from the small channel with/without the microporous coating were experimentally investigated. The small channel, even without the coating, presented flow boiling heat transfer enhancement at low vapor quality due to size effects of the channel. This enhancement was also verified by under-predictions from macro-scale correlations. In addition to the enhancement from the channel size, all six microporous coatings with various structural parameters were found to further enhance nucleate boiling significantly. Effects of the coating structural parameters, fluid mass flux and inlet subcooling were also investigated to identify the optimum condition for heat transfer enhancement. Under the optimum condition, the microporous coating could produce the heat transfer coefficients 2.7 times the smooth surface value in subcooled flow boiling and 3 times in saturated flow boiling. The combination of the microporous coating and small channel led to excellent heat transfer performance, and therefore was deemed to have promising application prospects in many areas such as air conditioning, chip cooling, refrigeration systems, and many others involving compact heat exchangers. (authors)

Cost estimation of hydrogen and DME produced by nuclear heat utilization system. Joint research

International Nuclear Information System (INIS)

Shiina, Yasuaki; Nishihara, Tetsuo

2003-09-01

Research of hydrogen energy has been performed in order to spread use of the hydrogen energy in 2020 or 2030. It will take, however, many years for the hydrogen energy to be used very easily like gasoline, diesel oil and city gas in all of countries. During the periods, low CO 2 release liquid fuels would be used together with hydrogen. Recently, di-methyl-either (DME) has been noticed as one of the substitute liquid fuels of petroleum. Such liquid fuels can be produced from the mixed gas such as hydrogen and carbon oxide which are produced by steam reforming hydrogen generation system by the use of nuclear heat. Therefore, the system would be one of the candidates of future system of nuclear heat utilization. In the present study, we focused on the production of hydrogen and DME. Economic evaluation was estimated for hydrogen and DME production in commercial and nuclear heat utilization plant. At first, heat and mass balance of each process in commercial plant of hydrogen production was estimated and commercial prices of each process were derived. Then, price was estimated when nuclear heat was used instead of required heat of commercial plant. Results showed that the production prices produced by nuclear heat were cheaper by 10% for hydrogen and 3% for DME. With the consideration of reduction effect of CO 2 release, utilization of nuclear heat would be more effective. (author)

Heat shock-induced interactions among nuclear HSFs detected by fluorescence cross-correlation spectroscopy

Energy Technology Data Exchange (ETDEWEB)

Pack, Chan-Gi, E-mail: changipack@amc.seoul.kr [Asan Institute for Life Sciences, University of Ulsan, College of Medicine, Asan Medical Center, Seoul 138-736 (Korea, Republic of); Ahn, Sang-Gun [Dept. of Pathology, College of Dentistry, Chosun University, Seosuk-dong, Dong-gu, Gwangju 501-759 (Korea, Republic of)

2015-07-31

The cellular response to stress is primarily controlled in cells via transcriptional activation by heat shock factor 1 (HSF1). HSF1 is well-known to form homotrimers for activation upon heat shock and subsequently bind to target DNAs, such as heat-shock elements, by forming stress granules. A previous study demonstrated that nuclear HSF1 and HSF2 molecules in live cells interacted with target DNAs on the stress granules. However, the process underlying the binding interactions of HSF family in cells upon heat shock remains unclear. This study demonstrate for the first time that the interaction kinetics among nuclear HSF1, HSF2, and HSF4 upon heat shock can be detected directly in live cells using dual color fluorescence cross-correlation spectroscopy (FCCS). FCCS analyses indicated that the binding between HSFs was dramatically changed by heat shock. Interestingly, the recovery kinetics of interaction between HSF1 molecules after heat shock could be represented by changes in the relative interaction amplitude and mobility. - Highlights: • The binding interactions among nuclear HSFs were successfully detected. • The binding kinetics between HSF1s during recovery was quantified. • HSF2 and HSF4 strongly formed hetero-complex, even before heat shock. • Nuclear HSF2 and HSF4 bound to HSF1 only after heat shock.

Study on a nuclear spaceship for interplanetary cruise. Core design of a small fast reactor

International Nuclear Information System (INIS)

Kitamura, Taku; Yoshida, Yutaka; Honma, Yuji; Narabayashi, Tadashi; Shimazu, Yoichiro; Tsuji, Masashi

2009-01-01

In 21st century, the field which needs nuclear power plant systems are not just on the Earth. We considered that the nuclear power is proper for the energy source of the manned spaceship for interplanetary cruise. In this study, we considered the system configuration of the spaceship, the design of power generating system, some navigational plans to reach the Mars. The system configuration of the spaceship studied in our laboratory has one or two Fast Reactor with liquid sodium coolant as main heat source, dozens of Stirling Engines as main power generators and some Plasma Rockets called VASIMR as propulsion system. Because the Fast Reactor need not thick and heavy pressure vessel and the sodium has high performance of heat transfer, they are the best suited to the space nuclear reactor system. In addition, Stirling Engine has high theoretical thermal efficiency and need not water, steam generators, steam condenser and so on. This results in absence of sodium-water reaction and significant weight saving of power generator system. The VASIMR studied at ASPL is an advanced electric propulsion device which is able to convert large amount of electric power into great propulsion force. At reactor designing, we are using the SRAC2006 code developed at JAEA and pursuing the optimal fast reactor design for spaceship. We think that smaller reactor is better. To realize a system which has inherent safety, sodium void reactivity should be negative. We adopted the design of the small fast reactor named 4S (Super Safe, Small and Simple) as a reference design. As a result, we verified that a void reactivity had negative value in some of calculation cases and we realized safe, small and simple space fast reactor. In addition, to piece out power generator system in space, we need to consider if the budget of exhaust heat from radiator panels to space needed at this case is realistic. To obtain the optimal trajectory of rapid Mars transit, we made some analysis calculation codes

Passive deca-heat removal in the fixed bed nuclear reactor (FBNR) - 15551

International Nuclear Information System (INIS)

Solano Diaz, E.C.; Luna Aguilera, G.M.; Santos, R.A.; Vaca, D.E.

2015-01-01

The Fixed Bed Nuclear Reactor (FBNR) is a Generation IV small reactor concept, where the spherical elements contain Triso-type microspheres with UO 2 , which serves as nuclear fuel. In the event that adverse operation conditions occur, the water pump is automatically shut off and the fuel pebbles fall back by gravity into the fuel chamber. Since the FBNR relies on passive security systems, the removal of the decay heat in the fuel chamber is achieved by contact with quiescent water. In the present paper, a mathematical simulation of the passive cooling of the system was conducted in SOLIDWORKS so as to obtain a temperature profile in the body during the decay heat removal process. Homogenization techniques were employed to smooth out spatial variations across the multiphase system and to derive expression for the effective thermophysical properties that are valid through the macroscopic entry (the chamber). The simulation showed that the chamber's temperature rose from 573 K to its maximum temperature, 1234 K, in the first hour. Afterwards, the temperature fluctuated, but stayed under 552 K. Since the temperature of the system was always kept under the value of the safety parameter (1200 C. degrees) the simulation confirmed that an effective passive cooling of the fuel chamber is indeed feasible. (authors)

Bibliographical survey of heat exchangers for nuclear power plants and problems of HTGR

International Nuclear Information System (INIS)

Yamao, Hiroyuki; Okamoto, Yoshizo; Sanokawa, Konomo

1977-04-01

The problems in development of heat exchangers for nuclear reactors have been examined in literature survey through Annual Index Subjects of NSA (Nuclear Science Abstracts) for the past ten years. R and D on heat exchangers for LMFBR, HTGR, LWR and HWR are on the increase. In the case of HTGRs, R and D on heat resisting materials including the corrosion and on hydrogen permeation of heat exchanger walls in high temperature pressure helium environment are important. Future R and D subjects for HTGR heat exchangers in showing the high temperature endurance are presented. (auth.)

An overview on the small heat shock proteins

African Journals Online (AJOL)

USER

2010-02-15

Feb 15, 2010 ... whose expression is increase when cells are exposed to elevated ... shock due to much slower degradation of the protein, .... Plant sHSPs are all encoded by nuclear genes and are .... genesis, germination, pollen growth and fruit maturation). ... Production of high levels of heat shock proteins can also.

Experimental simulation study on hydraulic behavior of the main heat exchanger of Daqing 200 MW nuclear heating reactor

International Nuclear Information System (INIS)

Jiang Shengyao; Zhang Youjie; Jia Haijun; Bo Jinhai; Hong Liuming; Bo Hanliang; Liu Zhiyong

1997-07-01

The hydraulic behavior of the main heat exchanger of Daqing 200 MW nuclear heating reactor is studied through a 1:2.33 test model. The design and other feature of the test model is described. The experimental results show that the flow resistance coefficient of the heat exchanger becomes self-simulation when Reynolds number is greater than 5000. The value of flow resistance coefficient at self-simulation condition and the distribution of pressure drop in the heat exchanger are given through experiment. The option design to reduce flow resistance is proposed. The designed and experimental value for the flow resistance coefficient are in good agreement. The variation of system parameters during flow excursion was described. The experimental results are of great significant for the final design of the main heat exchanger of Daqing 200 MW nuclear heating reactor. (2 refs., 5 figs., 1 tab.)

Heat Transfer Coefficient Variations in Nuclear Fuel Rod Bundles

International Nuclear Information System (INIS)

Conner, Michael E.; Holloway, Mary V.

2007-01-01

The single-phase heat transfer performance of a PWR nuclear fuel rod bundle is enhanced by the use of mixing vanes attached to the downstream edges of the support grid straps. This improved single-phase performance will delay the onset of nucleate boiling, thereby reducing corrosion and delaying crud-related issues. This paper presents the variation in measured single-phase heat transfer coefficients (HTC) for several grid designs. Then, this variation is compared with observations of actual in-core crud patterns. While crud deposition is a function of a number of parameters including rod heat flux, the HTC is assumed to be a primary factor in explaining why crud deposition is a local phenomenon on nuclear fuel rods. The data from this study will be used to examine this assumption by providing a comparison between HTC variations and crud deposition patterns. (authors)

Meeting Czechoslovak demands for heat in long-term prospective, especially with regard to nuclear sources

International Nuclear Information System (INIS)

Klail, M.

1988-01-01

The development was studied of heat demand in the CSSR till the year 2030. The ratio of centralized and decentralized heat supply is currently 60 to 40; in the future a slight increase is expected in the decentralized type of heat supply, mainly as a result of more intensive use of natural gas. In 2030, 710 PU of centralized heat should be produced. A decisive element in meeting the demand will be a growing proportion of combined production of electric power and heat by nuclear power plants. The installed capacity of the nuclear power plants in 2030 should range between 23 and 41 thousand MW, the production of electric power in these plants should be 193 to 238 TWh/y. 109 territorial areas potentially suitable for use of heat from nuclear sources were selected. They were included in 19 regions of which 9 should in the year 2010 be linked to heat supply from nuclear power plants that will be in operation. It is expected that in the year 2030, nuclear sources will supply 250 PU of centralized heat. (Z.M.). 2 tabs., 14 refs

Use of waste heat from nuclear power plants

International Nuclear Information System (INIS)

Olszewski, M.

1978-01-01

The paper details the Department of Energy (DOE) program concerning utilization of power plant reject heat conducted by the Oak Ridge National Laboratory (ORNL). A brief description of the historical development of the program is given and results of recent studies are outlined to indicate the scope of present efforts. A description of a DOE-sponsored project assessing uses for reject heat from the Vermont Yankee Nuclear Station is also given

Setting technical and economic features regarding nuclear heating plants implementation for heat supply in Romania by the year 2010

International Nuclear Information System (INIS)

Romascu, G.; Constantin, L.; Gheorghe, A.; Ciocanescu, M.; Ionescu, M.

2008-01-01

This paper presents the world wide preoccupation concerning the implementation of nuclear heating plants for fulfilling the heat demand and the main technical data of the reactors destined to such NHP's. The second part of this paper shows technical and economic aspects related to the implementation of NHP's equipped with nuclear thermal reactor specialized in the exclusive heat supply in Romania at the level of the year 2010. Among these aspects the following are mentioned: - the results of researches and the world wide achievements; - the development and structure of the production and of the thermal electric energy as well as the feasibility for covering the demands for nuclear sources; - the impact on environment of various technologies for the production of thermal energy with conventional fuels comparing with NHP; - the philosophy from economic stand point for the covering part of the NHP heat demand. (authors)

Fractionation of HeLa cell nuclear extracts reveals minor small nuclear ribonucleoprotein particles

International Nuclear Information System (INIS)

Kroemer, A.

1987-01-01

Upon chromatographic fractionation of HeLa cell nuclear extracts, small RNAs of 145 and 66/65 nucleotides, respectively, were detected that are distinct from the abundant small RNAs present in the extract. These RNAs are precipitated by antibodies directed against the trimethylguanosine cap structure, characteristic for small nuclear RNAs (snRNAs) of the U type. The RNAs of 145 and 66/65 nucleotides appear to be associated with at least one of the proteins common to the major small nuclear ribonucleoprotein particles U1 to U6, since they are specifically bound by anti-Sm antibodies. These criteria characterize the RNAs that are 145 and 66/65 nucleotides in length as U-type snRNAs. Upon gel filtration, the RNAs are found within particles of molecular weights ≅ 150,000 and 115,000 respectively. The RNA of 145 nucleotides represents a different minor snRNA, designated U11, whereas the RNA of 66/65 nucleotides may correspond to either mammalian U7 or U10 RNA

Time differentiated nuclear resonance spectroscopy coupled with pulsed laser heating in diamond anvil cells

Energy Technology Data Exchange (ETDEWEB)

Kupenko, I., E-mail: kupenko@esrf.fr; Strohm, C. [Bayerisches Geoinstitut, Universität Bayreuth, D-95440 Bayreuth (Germany); ESRF-The European Synchrotron, CS 40220, 38043 Grenoble Cedex 9 (France); McCammon, C.; Cerantola, V.; Petitgirard, S.; Dubrovinsky, L. [Bayerisches Geoinstitut, Universität Bayreuth, D-95440 Bayreuth (Germany); Glazyrin, K. [Photon Science, DESY, D-22607 Hamburg (Germany); Vasiukov, D.; Aprilis, G. [Laboratory of Crystallography, Material Physics and Technology at Extreme Conditions, Universität Bayreuth, D-95440 Bayreuth (Germany); Chumakov, A. I.; Rüffer, R. [ESRF-The European Synchrotron, CS 40220, 38043 Grenoble Cedex 9 (France)

2015-11-15

Developments in pulsed laser heating applied to nuclear resonance techniques are presented together with their applications to studies of geophysically relevant materials. Continuous laser heating in diamond anvil cells is a widely used method to generate extreme temperatures at static high pressure conditions in order to study the structure and properties of materials found in deep planetary interiors. The pulsed laser heating technique has advantages over continuous heating, including prevention of the spreading of heated sample and/or the pressure medium and, thus, a better stability of the heating process. Time differentiated data acquisition coupled with pulsed laser heating in diamond anvil cells was successfully tested at the Nuclear Resonance beamline (ID18) of the European Synchrotron Radiation Facility. We show examples applying the method to investigation of an assemblage containing ε-Fe, FeO, and Fe{sub 3}C using synchrotron Mössbauer source spectroscopy, FeCO{sub 3} using nuclear inelastic scattering, and Fe{sub 2}O{sub 3} using nuclear forward scattering. These examples demonstrate the applicability of pulsed laser heating in diamond anvil cells to spectroscopic techniques with long data acquisition times, because it enables stable pulsed heating with data collection at specific time intervals that are synchronized with laser pulses.

Small Nuclear Technology and Market Entry

International Nuclear Information System (INIS)

Stewart, J S; Schock, R N; Brown, N W; Smith, C F

2002-01-01

An overview of energy-system projections into the new century leads to the conclusion that nuclear power will play a significant role. How significant a role will be determined by the marketplace. Within the range of nuclear-power technologies available, small nuclear-power plants of innovative design appear to fit the needs of a number of developing nations and states. Under similar financing options used by the airline industry and others, the capital requirement barrier that puts the nuclear industry at a disadvantage in deregulated markets could be reduced. These plants have the potential advantage of modularity, are proliferation-resistant, incorporate passive safety features, minimize waste, and could be cost-competitive with fossil-fuel plants

Heat transfer and mechanical interactions in fusion nuclear systems

International Nuclear Information System (INIS)

Nygren, R.E.

1984-01-01

This general review of design issues in heat transfer and mechanical interactions of the first wall, blanket and shield systems of tokamak and mirror fusion reactors begins with a brief introduction to fusion nuclear systems. The design issues are summarized in tables and the following examples are described to illustrate these concerns: the surface heating of limiters, heat transfer from solid breeders, MHD effects in liquid metal blankets, mechanical loads from electromagnetic transients and remote maintenance

Multi-unit Operations in Non-Nuclear Systems: Lessons Learned for Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

OHara J. M.; Higgins, J.; DAgostino, A.

2012-01-17

The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. Small modular reactors (SMRs) are one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants and may be operated quite differently. One difference is that multiple units may be operated by a single crew (or a single operator) from one control room. The U.S. Nuclear Regulatory Commission (NRC) is examining the human factors engineering (HFE) aspects of SMRs to support licensing reviews. While we reviewed information on SMR designs to obtain information, the designs are not completed and all of the design and operational information is not yet available. Nor is there information on multi-unit operations as envisioned for SMRs available in operating experience. Thus, to gain a better understanding of multi-unit operations we sought the lesson learned from non-nuclear systems that have experience in multi-unit operations, specifically refineries, unmanned aerial vehicles and tele-intensive care units. In this paper we report the lessons learned from these systems and the implications for SMRs.

Design guide for heat transfer equipment in water-cooled nuclear reactor systems

International Nuclear Information System (INIS)

1975-07-01

Information pertaining to design methods, material selection, fabrication, quality assurance, and performance tests for heat transfer equipment in water-cooled nuclear reactor systems is given in this design guide. This information is intended to assist those concerned with the design, specification, and evaluation of heat transfer equipment for nuclear service and the systems in which this equipment is required. (U.S.)

Salt disposal of heat-generating nuclear waste

International Nuclear Information System (INIS)

Leigh, Christi D.; Hansen, Francis D.

2011-01-01

This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from United

Salt disposal of heat-generating nuclear waste.

Energy Technology Data Exchange (ETDEWEB)

Leigh, Christi D. (Sandia National Laboratories, Carlsbad, NM); Hansen, Francis D.

2011-01-01

This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from

Utilization of waste heat from nuclear power plants in agriculture

International Nuclear Information System (INIS)

Horacek, P.

1981-01-01

The development of nuclear power will result in the relative and absolute increase in the amount of waste heat which can be used in agriculture for heating greenhouses, open spaces, for fish breeding in heated water, for growing edible mushrooms, growing algae, for frost protection of orchards, air conditioning of buildings for breeding livestock and poultry, and for other purposes. In addition of the positive effect of waste heat, the danger increases of disease, weeds and pests. Pilot plant installations should be build in Czechoslovakia for testing the development of waste heat utilization. (Ha)

The Sydvaerme project: District heating from the Barsebeck nuclear power plant

International Nuclear Information System (INIS)

Josefsson, L.

1977-01-01

The paper presents a summary report of a study on district heating from Barsebeck Nuclear Power Plant in Sweden, prepared cooperatively by the cities of Malmoe, Lund, Helsingborg, Landskrona and the electric power company Sydkraft. A future number 3 generating set at the Barsebeck nuclear power station could be designed for combined production of heat and electric power. The generating set could be completed after 1983, and could then supply about 65% of total district heating requirements. The first stage of the investigation includes a proposal for a technically feasible solution, sufficiently detailed to permit both technical and economic evaluation of the project. (author)

Evaporation and condensation devices for passive heat removal systems in nuclear power engineering

International Nuclear Information System (INIS)

Gershuni, A.N.; Pis'mennyj, E.N.; Nishchik, A.P.

2016-01-01

The paper justifies advantages of evaporation and condensation heat transfer devices as means of passive heat removal and thermal shielding in nuclear power engineering. The main thermophysical factors that limit heat transfer capacity of evaporation and condensation systems have been examined in the research. The results of experimental studies of heat engineering properties of elongated (8-m) vertically oriented evaporation and condensation devices (two-phase thermosyphons), which showed a high enough heat transfer capacity, as well as stability and reliability both in steady state and in start-up modes, are provided. The paper presents the examples of schematic designs of evaporation and condensation systems for passive heat removal and thermal shielding in application to nuclear power equipment

Comparative analysis of heat pump and biomass boiler for small detached house heating

Directory of Open Access Journals (Sweden)

Olkowski Tomasz

2017-01-01

Full Text Available The purpose of the work is to answer the question - which of the two selected heat sources is more economically beneficial for small detached house: heat pump or biomass boiler fuelled with wood-pellets? The comparative analysis of these sources was carried out to discuss the issue. First, cost of both, equipment and operation of selected heat systems were analysed. Additionally, CO2 emission levels associated with these heat systems were determined. The comparative analysis of the costs of both considered heat systems showed that equipment cost of heat pump system is considerably bigger than the cost of biomass boiler system. The comparison of annual operation costs showed that heat pump operation cost is slightly lower than operation cost of biomass boiler. The analysis of above results shows that lower operation cost of heat pump in comparison with biomass boiler cost lets qualify heat pump as more economically justified only after 38 years of work. For both analysed devices, CO2 emission levels were determined. The considerations take into account the fact that heat pump consumes electricity. It is mostly generated through combustion of coal in Poland. The results show that in Poland biomass boiler can be described as not only more economically justified system but also as considerably more ecological.

Comparative analysis of heat pump and biomass boiler for small detached house heating

Science.gov (United States)

Olkowski, Tomasz; Lipiński, Seweryn; Olędzka, Aneta

2017-10-01

The purpose of the work is to answer the question - which of the two selected heat sources is more economically beneficial for small detached house: heat pump or biomass boiler fuelled with wood-pellets? The comparative analysis of these sources was carried out to discuss the issue. First, cost of both, equipment and operation of selected heat systems were analysed. Additionally, CO2 emission levels associated with these heat systems were determined. The comparative analysis of the costs of both considered heat systems showed that equipment cost of heat pump system is considerably bigger than the cost of biomass boiler system. The comparison of annual operation costs showed that heat pump operation cost is slightly lower than operation cost of biomass boiler. The analysis of above results shows that lower operation cost of heat pump in comparison with biomass boiler cost lets qualify heat pump as more economically justified only after 38 years of work. For both analysed devices, CO2 emission levels were determined. The considerations take into account the fact that heat pump consumes electricity. It is mostly generated through combustion of coal in Poland. The results show that in Poland biomass boiler can be described as not only more economically justified system but also as considerably more ecological.

Valorization of the energy potential of fossil and fissile fuels for heat production: dual-purpose power plants and heat-producing nuclear reactors

International Nuclear Information System (INIS)

Lavite, Michel.

1975-07-01

The heat market is analyzed briefly within the French context: present structures and characteristics of the market, current means of heat production, predictable trend of the demand. The possible applications of nuclear energy to heat production, through the agency of combined electricity-steam stations or heat-producing stations, are then examined. Nuclear solutions are compared with others from the technico-economic and ecological wiewpoints and an estimate fo their respective impacts on the energy balance is attempted [fr

Nuclear heat generating plants - technical concepts and market potentials. Chapter 11

International Nuclear Information System (INIS)

Hasenkopf, O.; Erhard, W.D.; Nonnenmacher, A.; Hanselmann, M.

1988-01-01

Within the framework of a case study under the Federal Ministry of Research and Technology project 'Nuclear heat generating plants - technological concepts and market potentials', the possible applications of such plants were studied giving the district heat supply network of the Technische Werke der Stadt Stuttgart AG (Technical Works of the City of Stuttgart, Inc.) as an example. The use of district heating systems concentrated specifically on areas identified for economical supply because of their topographical position, existing heat density, distance from power plants, and a reasonable delimination from the available gas network. Based on the results of optimization calculations made by the Stuttgart Institute for Nuclear Technology and Energy Conversion, the required investment capital can be estimated as a function of the amount of fuel savings under the Stuttgart case study. (orig./UA) [de

A small floating seawater desalination plant by using a nuclear heating reactor coupled with the MED process

Energy Technology Data Exchange (ETDEWEB)

Xue Dazhi; Zhang Dafang; Dong Duo [Institute of Nuclear Energy Technology, Tsinghua University, Beijing (China)

2000-03-01

Based on the experience of development of nuclear district heating reactor (NHR) a seawater desalination plant using NHR coupled with the multi-effect distillation (MED) process is being designed. With the same technology a floating desalination plant was proposed to supply potable water to remote areas or islands. With a 10 MWth NHR the floating plant could produce 4000 m{sup 3}/d of potable water and 750 kW of electricity. The design of NHR-10 and the safety features are described. The coupling scheme and parameters are given. Some special considerations for using in ship condition are also presented in this paper. (author)

Qualification of γ-heating calculation in nuclear reactors

International Nuclear Information System (INIS)

Ravaux, Simon

2013-01-01

During the last few years, the γ-heating issue has gained in stature, mainly for the safety of the 3. generation reactors in which a stainless steel reflector is inserted. The purpose of this work is the qualification of the needed tools for calculation of the γ-heating in the nuclear reactors. In a nuclear reactor, all the photons are directly or indirectly produced by the neutron-matter interactions. Thus, the first phase of this work is a critical analysis of the photon production data in the standard nuclear data library. New evaluations have been proposed to the next version of the JEFF library after that some omissions have been found. They have partly been accepted for JEFF-3.2. Two particle-transport codes are currently developed in the CEA: the deterministic code APOLLO2 and the Monte Carlo code TRIPOLI4. The second part of this work is the qualification of both these codes by interpreting an integral experiment called PERLE. The experimental set-up is made by a LWR pin assembly surrounded by a stainless steel reflector in which the γ-heating is measured by Thermo-luminescent Detector (TLD). A calculation scheme has been proposed for both APOLLO2 and TRIPOLI4 in order to calculate the TLD's responses. Comparisons between calculations and measurements have shown that TRIPOLI4 gives a satisfactory estimation of the γ-heating in the reflector. These discrepancies are within the experimental 1 σ uncertainty. Before the qualification, APOLLO2 has been previously validated against TRIPOLI4 reference calculation. This validation gives an estimation of the bias due to the deterministic approximations of the transport equation resolution. The qualification has shown that the discrepancies between APOLLO2 predictions and TLD's measurements are in the same range as experimental uncertainties. (author) [fr

Preliminary study on high temperature heat exchanger for nuclear steel making

Energy Technology Data Exchange (ETDEWEB)

Nakada, T; Ohtomo, A; Yamada, R; Suzuki, K; Narita, Y [Ishikawajima-Harima Heavy Industries Co. Ltd., Tokyo (Japan)

1975-05-01

Both in the high temperature heat exchanger and in the steam reformer, there remain several technical problems to be solved before nuclear steel making is actualized. The loop for use with basic studies of those problems was planned by the Iron and Steel Institute of Japan (ISIJ), and its actual design, construction and co-ordination of tests were undertaken by IHI on behalf of ISIJ. The primary coolant used in the loop was helium having a pressure of approx. 12 kg/cm/sup 2/g and a temperature of approx. 1100/sup 0/C at the inlet of the high temperature heat exchanger, i.e., the test section. Steam, hydrogen, and carbon monoxide were used as secondary coolants. Of the technical problems regarding the high temperature heat exchanger for nuclear steel making, which were selected and studied using the loop, the following items are discussed: (1) heat exchange performance using helium and steam; (2) hydrogen permeation of heat resisting alloys; (3) creep and carburization of heat resisting alloys; amd (4) hydrogen absorption performance of the titanium sponge.

Economic Analyses and Potential Market of the 200MW Nuclear Heating Reactor

International Nuclear Information System (INIS)

Wang, Yongqing; Wang, Guiying

1992-01-01

Based on the 5MW experimental nuclear heating reactor, Intent has developed a 200MW demonstration nuclear heating reactor. Owing to its simplified systems and low operating parameters, the NCR-200 has preferable investment in comparison with that of a nuclear power plant. The pre-feasibility studies for several cities in Northern China have shown that the heat cost of a NCR-200 can be competitive with a coal fired heating plant. As a safe, clean and economic heat source, the NCR could pose a large market in replacement of coal for heating. The R and D work performed up to now has demonstrated that the NCR-200 operating under the present parameters can supply low pressure steam for industrial process and co-generation to enhance it economic benefit. The NCR-200 could also serve a heat source for air condition by using Li Br refrigerator, this application is very interesting to some cities in Central and Southern China. The applications of the NCR in oil recovery by injecting hot water and transportation are very promising for some oil fields in North China. In addition, the study on sea water desalination using the NCR-200 is being carried out at present under international cooperation. All of these will expansion the possible application of the NCR. The paper presents the economic analysis and the potential market of the NCR-200

Westinghouse Small Modular Reactor nuclear steam supply system design

Energy Technology Data Exchange (ETDEWEB)

Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

Cost estimation of hydrogen and DME produced by nuclear heat utilization system II

International Nuclear Information System (INIS)

Shiina, Yasuaki; Nishihara, Tetsuo

2004-09-01

Utilization and production of hydrogen has been studied in order to spread utilization of the hydrogen energy in 2020 or 2030. It will take, however, many years for the hydrogen energy to be used very easily like gasoline, diesel oil and city gas in the world. During the periods, low CO 2 release liquid fuels would be used together with hydrogen. Recently, di-methyl-ether (DME). has been noticed as one of the substitute liquid fuels of petroleum. Such liquid fuels can be produced from the mixed gas such as hydrogen and carbon oxide which are produced from natural gas by steam reforming. Therefore, the system would become one of the candidates of future system of nuclear heat utilization. Following the study in 2002, we performed economic evaluation of the hydrogen and DME production by nuclear heat utilization plant where heat generated by HTGR is completely consumed for the production. The results show that hydrogen price produced by nuclear was about 17% cheaper than the commercial price by increase in recovery rate of high purity hydrogen with increased in PSA process. Price of DME in indirect method produced by nuclear heat was also about 17% cheaper than the commercial price by producing high purity hydrogen in the DME producing process. As for the DME, since price of DME produced near oil land in petroleum exporting countries is cheaper than production in Japan, production of DME by nuclear heat in Japan has disadvantage economically in this time. Trial study to estimate DME price produced by direct method was performed. From the present estimation, utilization of nuclear heat for the production of hydrogen would be more effective with coupled consideration of reduction effect of CO 2 release. (author)

Next Generation Nuclear Plant Intermediate Heat Exchanger Acquisition Strategy

Energy Technology Data Exchange (ETDEWEB)

Mizia, Ronald Eugene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2008-04-01

DOE has selected the High Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C to 950°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium cooled, prismatic or pebble-bed reactor, and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. The purpose of this report is to address the acquisition strategy for the NGNP Intermediate Heat Exchanger (IHX).This component will be operated in flowing, impure helium on the primary and secondary side at temperatures up to 950°C. There are major high temperature design, materials availability, and fabrication issues that need to be addressed. The prospective materials are Alloys 617, 230, 800H and X, with Alloy 617 being the leading candidate for the use at 950°C. The material delivery schedule for these materials does not pose a problem for a 2018 start up as the vendors can quote reasonable delivery times at the moment. The product forms and amount needed must be finalized as soon as possible. An

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

Heat-electricity convertion systems for a Brazilian space micro nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Guimaraes, Lamartine N.F.; Marcelino, Natalia B.; Placco, Guilherme M.; Nascimento, Jamil A.; Borges, Eduardo M., E-mail: guimarae@ieav.cta.br, E-mail: lamartine.guimaraes@pq.cnpq.br, E-mail: jamil@ieav.cta.br, E-mail: jalnsgf@outlook.com, E-mail: borges.em@hotmail.com, E-mail: ecorborges@hotmail.com, E-mail: ivayolini@gmail.com, E-mail: guilherme_placco@ig.com.br [Instituto de Estudos Avancados (IEAv/DCTA), Sao Jose dos Campos, SP (Brazil); Barrios Junior, Ary Garcia, E-mail: arygarcia89@yahoo.com [Faculdade de Tecnologia Sao Francisco (FATESF), Jacarei, SP (Brazil)

2013-07-01

This contribution will discuss the evolution work in the development of thermal cycles to allow the development of heat-electricity conversion for the Brazilian space micro nuclear Reactor. Namely, innovative core and nuclear fuel elements, Brayton cycle, Stirling engine, heat pipes, passive multi-fluid turbine, among others. This work is basically to set up the experimental labs that will allow the specification and design of the space equipment. Also, some discussion of the cost so far, and possible other applications will be presented. (author)

Heat-electricity convertion systems for a Brazilian space micro nuclear reactor

International Nuclear Information System (INIS)

Guimaraes, Lamartine N.F.; Marcelino, Natalia B.; Placco, Guilherme M.; Nascimento, Jamil A.; Borges, Eduardo M.; Barrios Junior, Ary Garcia

2013-01-01

This contribution will discuss the evolution work in the development of thermal cycles to allow the development of heat-electricity conversion for the Brazilian space micro nuclear Reactor. Namely, innovative core and nuclear fuel elements, Brayton cycle, Stirling engine, heat pipes, passive multi-fluid turbine, among others. This work is basically to set up the experimental labs that will allow the specification and design of the space equipment. Also, some discussion of the cost so far, and possible other applications will be presented. (author)

Numerical simulation of flow field in shellside of heat exchanger in nuclear power plant

International Nuclear Information System (INIS)

Wang Xinliang; Qiu Jinrong; Gong Zili

2010-01-01

Heat exchanger is the important equipment of nuclear power plant. Numerical simulation can give the detail information inside the heat exchange, and has been an effective research method. The geometric structure of shell-and-tube heat exchanger is very complex and it is difficult to simulate the whole flow field presently. According to the structure characteristics of the heat exchanger, a periodic whole-section calculation model was presented. The numerical simulation of flow field in shellside of heat exchange of a nuclear power plant was done by using this model. The results of simulation show that heat transfer in the periodic section of the heat exchange is uniform, the heat transfer is enhanced by using baffles in heat exchange, and frictional resistance is primary from the effect of segmental baffles. (authors)

Nuclear power generation and global heating

International Nuclear Information System (INIS)

Taboada, Horacio

1999-01-01

The Professionals Association and Nuclear Activity of National Atomic Energy Commission (CNEA) are following with great interest the worldwide discussions on global heating and the role that nuclear power is going to play. The Association has an active presence, as part of the WONUC (recognized by the United Nations as a Non-Governmental Organization) in the COP4, which was held in Buenos Aires in November 1998. The environmental problems are closely related to human development, the way of power production, the techniques for industrial production and exploitation fields. CO 2 is the most important gas with hothouse effects, responsible of progressive climatic changes, as floods, desertification, increase of average global temperature, thermal expansion in seas and even polar casks melting and ice falls. The consequences that global heating will have on the life and economy of human society cannot be sufficiently emphasized, great economical impact, destruction of ecosystems, loss of great coast areas and complete disappearance of islands owing to water level rise. The increase of power retained in the atmosphere generates more violent hurricanes and storms. In this work, the topics presented in the former AATN Meeting is analyzed in detail and different technological options and perspectives to mitigate CO 2 emission, as well as economical-financial aspects, are explored. (author)

Regional energy-environment system analysis and the role of low-temperature nuclear heat in North China

International Nuclear Information System (INIS)

Lu Yingyun

1984-01-01

The consumption of commercial energy in China in 1980 amounted to 603 million tonnes of coal equivalent (tce). By the end of this century, according to preliminary forecasting, it will reach some 1200 million tce at least, but there may still be some gaps in the energy supply. Within the structure of China's current energy supply, coal is the dominating fuel, most of which is burned directly, thus causing serious air pollution particularly in urban areas during the winter season. To take into consideration the environmental impacts in formulating appropriate energy policies and carrying out rational energy planning, a practical regional energy system model in connection with environment impacts has been developed. It is essentially a linear programme model. The model has already been used to evaluate the role of alternative energies and technologies including the nuclear option in North China's future urban energy system. The preliminary results thus obtained have shown that nuclear energy, particularly low-temperature nuclear heat, must be introduced to reduce air pollution and fill the gaps in the energy supply. Since small- or medium-sized heat-only reactors have already been reported to be economical, safe and non-polluting, that will be developed in urban areas in North China to a certain extent by the end of this century. (author)

Nuclear heat applications: design aspects and operating experience. Proceedings of four technical meetings held between December 1995 and April 1998

International Nuclear Information System (INIS)

1998-11-01

Proven to be safe, reliable, economic and having minimum impact on the environment, nuclear energy is playing an important role in electricity generation producing 175 of the world's electricity. But since most of the world's energy consumption is in the form of heat the market for nuclear heat has already been recognised. Considering the growing experience in application of power reactors for district heating, industrial processes and water desalination IAEA is periodically reviewing progress and new developments of nuclear heat applications. This proceedings includes the papers presented at the following four meetings: Advisory group meeting and consultancy on experience with nuclear heat applications: district heating, process heat and desalination, 13-15 December 1995 and 7-9 february 1996; Advisory group meeting on technology, design and safety aspects of non-electrical application of nuclear energy, 20-24 October 1997; Advisory group meeting on operational modes of nuclear desalination plants, 3-5 November 1997; Advisory group meeting on materials and equipment for the coupling interfaces of nuclear reactors with desalination and district heating plants, 21-23 April 1998. It is structured according to the subject areas: (1) design an safety aspects of nuclear heat application, (2) operational and material aspects of nuclear heat application and (3) operational experience with nuclear heat application. Each paper is described by a separate abstract

Small-scale heat detection using catalytic microengines irradiated by laser

Science.gov (United States)

Liu, Zhaoqian; Li, Jinxing; Wang, Jiao; Huang, Gaoshan; Liu, Ran; Mei, Yongfeng

2013-01-01

We demonstrate a novel approach to modulating the motion speed of catalytic microtubular engines via laser irradiation/heating with regard to small-scale heat detection. Laser irradiation on the engines leads to a thermal heating effect and thus enhances the engine speed. During a laser on/off period, the motion behaviour of a microengine can be repeatable and reversible, demonstrating a regulation of motion speeds triggered by laser illumination. Also, the engine velocity exhibits a linear dependence on laser power in various fuel concentrations, which implies an application potential as local heat sensors. Our work may hold great promise in applications such as lab on a chip, micro/nano factories, and environmental detection.We demonstrate a novel approach to modulating the motion speed of catalytic microtubular engines via laser irradiation/heating with regard to small-scale heat detection. Laser irradiation on the engines leads to a thermal heating effect and thus enhances the engine speed. During a laser on/off period, the motion behaviour of a microengine can be repeatable and reversible, demonstrating a regulation of motion speeds triggered by laser illumination. Also, the engine velocity exhibits a linear dependence on laser power in various fuel concentrations, which implies an application potential as local heat sensors. Our work may hold great promise in applications such as lab on a chip, micro/nano factories, and environmental detection. Electronic supplementary information (ESI) available. See DOI: 10.1039/c2nr32494f

Peltier heat of a small polaron in a magnetic semiconductor

International Nuclear Information System (INIS)

Liu, N.H.; Emin, D.

1985-01-01

For the first time the heat transported with a small polaron in both antiferromagnetic and ferromagnetic semiconductors is calculated. This heat, the Peltier heat, π, is obtained from the change of the entropy of the total system upon introduction of a charge carrier. We explicitly consider both the intrasite and intersite exchange interactions between a small polaron and the interacting spins of a spin-1/2 magnet. There are two competing magnetic contributions to the Peltier heat. First, adding the carrier increases the spin entropy of the system. This provides a positive contribution to π. Second, the exchange between the carrier and the sites about it enhances the exchange binding between these sites. This reduces the energetically allowable spin configurations and provides a negative contribution to π. At extremely high temperatures when kT exceeds the intrasite exchange energy, the first effect dominates. Then π is simply augmented by kT ln 2. However, well below the magnetic transition temperature the second effect dominates. In the experimentally accessible range between these limits both effects are comparable and sizable. The net magnetic contribution to the Peltier heat rises with temperature. Thus, a carrier's interactions with its magnetic environment produces a significant and distinctive contribution to its Peltier heat

Peltier heat of a small polaron in a magnetic semiconductor

International Nuclear Information System (INIS)

Liu, N.L.H.; Emin, D.

1984-01-01

The heat transported with a small polaron in both antiferromagnetic and ferromagnetic semiconductors is calculated. This heat, the Peltier heat, π, is obtained from the change of the entropy of the total system upon introduction of a charge carrier. We explicitly consider both the intrasite and intersite exchange interactions between a small polaron and the interacting spins of a spin-1/2 magnet. There are two competing magnetic contributions to the Peltier heat. First, adding the carrier increases the spin entropy of the system. This provides a positive contribution to π. Second, the exchange between the carrier and the sites about it enhances the exchange binding between these sites. This reduces the energetically allowable spin configurations and provides a negative contribution to π. At extremely high temperature when kT exceeds the intrasite exchange energy, the first effect dominates. Then π is simply augmented by kTln2. However, well below the magnetic transition temperature the second effect dominates. In the experimentally accessible range between these limits both effects are comparable and sizable. The net magnetic contribution to the Peltier heat rises with temperature. Thus, a carrier's interactions with its magnetic environment produces a significant and distinctive contribution to its Peltier heat

IAEA Expands International Cooperation on Small, Medium Sized or Modular Nuclear Reactors

International Nuclear Information System (INIS)

2018-01-01

The International Atomic Energy Agency (IAEA) is launching an effort to expand international cooperation and coordination in the design, development and deployment of small, medium sized or modular reactors (SMRs), among the most promising emerging technologies in nuclear power. Significant advances have been made on SMRs, some of which will use pre-fabricated systems and components to shorten construction schedules and offer greater flexibility and affordability than traditional nuclear power plants. With some 50 SMR concepts at various stages of development around the world, the IAEA is forming a Technical Working Group (TWG) to guide its activities on SMRs and provide a forum for Member States to share information and knowledge, IAEA Deputy Director General Mikhail Chudakov said. “Innovation is crucial for nuclear power to play a key role in decarbonising the energy sector,” Chudakov, who heads the IAEA Department of Nuclear Energy, said at a conference on SMRs in Prague on 15 February. “Many Member States that are operating, expanding, introducing or considering nuclear power are quite keen on the development and deployment of SMRs.” Global interest in SMRs is growing. SMRs have the potential to meet the needs of a wide range of users and to be low carbon replacements for ageing fossil fuel fired power plants. They also display enhanced safety features and are suitable for non-electric applications, such as cooling, heating and water desalination. In addition, SMRs offer options for remote regions with less developed infrastructure and for energy systems that combine nuclear and alternative sources, including renewables.

Synergistic production of hydrogen using fossil fuels and nuclear energy application of nuclear-heated membrane reformer

International Nuclear Information System (INIS)

Hori, M.; Matsui, K.; Tashimo, M.; Yasuda, I.

2004-01-01

Processes and technologies to produce hydrogen synergistically by the steam reforming reaction using fossil fuels and nuclear heat are reviewed. Formulas of chemical reactions, required heats for reactions, saving of fuel consumption or reduction of carbon dioxide emission, possible processes and other prospects are examined for such fossil fuels as natural gas, petroleum and coal. The 'membrane reformer' steam reforming with recirculation of reaction products in a closed loop configuration is considered to be the most advantageous among various synergistic hydrogen production methods. Typical merits of this method are: nuclear heat supply at medium temperature below 600 deg. C, compact plant size and membrane area for hydrogen production, efficient conversion of feed fuel, appreciable reduction of carbon dioxide emission, high purity hydrogen without any additional process, and ease of separating carbon dioxide for future sequestration requirements. With all these benefits, the synergistic production of hydrogen by membrane reformer using fossil fuels and nuclear energy can be an effective solution in this century for the world which has to use. fossil fuels any way to some extent while reducing carbon dioxide emission. For both the fossil fuels industry and the nuclear industry, which are under constraint of resource, environment and economy, this production method will be a viable symbiosis strategy for the coming hydrogen economy era. (author)

Transwaal - economic district heat from the Beznau nuclear power station

International Nuclear Information System (INIS)

Schatzmann, G.

1986-01-01

Initial study phases of the Transwaal project for distribution of heat from the Beznau nuclear power station via pipe lines to Aare and Limmat valley regions in Switzerland are presented. 500 MW heat availability through heat exchangers providing forward flow water temperature of 120 0 C, pipe line network and pumping station aspects, and the system energy flow diagram, are described. Considerations based on specific energy requirements in the year 2000 including alternative schemes showed economic viability. Investment and consumer costs and savings compared with oil and gas heating are discussed. Heat supply is guaranteed well into the 21st century and avoids environmental disadvantages. (H.V.H.)

Steam turbines for nuclear power stations in Czechoslovakia and their use for district heating

International Nuclear Information System (INIS)

Drahy, J.

1989-01-01

The first generation of nuclear power stations in Czechoslavakia is equipped with 440 MW e pressurized water reactors. Each reactor supplies two 220 MW, 3000 rpm condensing type turbosets operating with saturated steam. After the completion of heating water piping systems, all of the 24 units of 220 MW in Czechoslovak nuclear power stations will be operated as dual purpose units, delivering both electricity and heat. At the present time, second-generation nuclear power stations, with 1000 MW e PWRs, are being built. Each such plant is equipped with one 1000 MW full-speed saturated steam turbine. The turbine is so designed as to permit the extraction of steam corresponding to the following quantities of heat: 893 MJ/s with three-stage water heating (150/60 0 C); and 570 MJ/s with two-stage water heating (120/60 0 C). The steam is taken from uncontrolled steam extraction points. (author)

Thermal hydraulic analysis of the encapsulated nuclear heat source

Energy Technology Data Exchange (ETDEWEB)

Sienicki, J.J.; Wade, D.C. [Argonne National Lab., IL (United States)

2001-07-01

An analysis has been carried out of the steady state thermal hydraulic performance of the Encapsulated Nuclear Heat Source (ENHS) 125 MWt, heavy liquid metal coolant (HLMC) reactor concept at nominal operating power and shutdown decay heat levels. The analysis includes the development and application of correlation-type analytical solutions based upon first principles modeling of the ENHS concept that encompass both pure as well as gas injection augmented natural circulation conditions, and primary-to-intermediate coolant heat transfer. The results indicate that natural circulation of the primary coolant is effective in removing heat from the core and transferring it to the intermediate coolant without the attainment of excessive coolant temperatures. (authors)

Quercetin suppresses heat shock-induced nuclear translocation of Hsp72

Directory of Open Access Journals (Sweden)

Antoni Gawron

2011-08-01

Full Text Available The effect of quercetin and heat shock on the Hsp72 level and distribution in HeLa cells was studied by Western blotting, indirect immunofluorescence and immunogold electron microscopy. In control cells and after quercetin treatment, Hsp72 was located both in the cytoplasm and in the nucleus in comparable amounts. After hyperthermia, the level of nuclear Hsp72 raised dramatically. Expression of Hsp72 in cytoplasm was also higher but not to such extent as that observed in the nucleus. Preincubation of heated cells with quercetin inhibited strong Hsp72 expression observed after hyperthermia and changed the intracellular Hsp72 distribution. The cytoplasmic level of protein exceeded the nuclear one, especially around the nucleus, where the coat of Hsp72 was noticed. Observations indicating that quercetin was present around and in the nuclear envelope suggested an involvement of this drug in the inhibition of nuclear translocation. Our results indicate that pro-apoptotic activity of quercetin may be correlated not only with the inhibition of Hsp72 expression but also with suppression of its migration to the nucleus.

Cooling and heating facility for nuclear power plant

International Nuclear Information System (INIS)

Kakuta, Atsuro

1994-01-01

The present invention concerns a cooling and heating facility for a nuclear power plant. Namely, a cooling water supply system supplies cooling water prepared by a refrigerator for cooling the inside of the plant. A warm water supply system supplies warm water having its temperature elevated by using an exhausted heat from a reactor water cleanup system. The facility comprises a heat pump-type refrigerator disposed in a cold water supply system for producing cold water and warm water, and warm water pipelines for connecting the refrigerator and the warm water supply system. With such a constitution, when the exhaust heat from the reactor water cleanup system can not be used, warm water prepared by the heat pump type refrigerator is supplied to the warm water supply system by way of the warm water pipelines. Accordingly, when the exhaust heat from the reactor water cleanup system can not be used such as upon inspection of the plant, a portion of the refrigerators in a not-operated state can be used for heating. Supply of boiler steams in the plant is no more necessary or extremely reduced. (I.S.)

Heat and mass transfer and hydrodynamics in two-phase flows in nuclear power plants

International Nuclear Information System (INIS)

Styrikovich, M.A.; Polonskii, V.S.; Tsiklauri, G.V.

1986-01-01

This book examines nuclear power plant equipment from the point of view of heat and mass transfer and the behavior of impurities contained in water and in steam, with reference to real water regimes of nuclear power plants. The transfer processes of equipment are considered. Heat and mass transfer are analyzed in the pre-crisis regions of steam-generating passages with non-permeable surfaces, and in capillary-porous structures. Attention is given to forced convection boiling crises and top post-DNB heat transfer. Data on two-phase hydrodynamics in straight and curved channels are correlated and safety aspects of nuclear power plants are discussed

Laser-heating and Radiance Spectrometry for the Study of Nuclear Materials in Conditions Simulating a Nuclear Power Plant Accident.

Science.gov (United States)

Manara, Dario; Soldi, Luca; Mastromarino, Sara; Boboridis, Kostantinos; Robba, Davide; Vlahovic, Luka; Konings, Rudy

2017-12-14

Major and severe accidents have occurred three times in nuclear power plants (NPPs), at Three Mile Island (USA, 1979), Chernobyl (former USSR, 1986) and Fukushima (Japan, 2011). Research on the causes, dynamics, and consequences of these mishaps has been performed in a few laboratories worldwide in the last three decades. Common goals of such research activities are: the prevention of these kinds of accidents, both in existing and potential new nuclear power plants; the minimization of their eventual consequences; and ultimately, a full understanding of the real risks connected with NPPs. At the European Commission Joint Research Centre's Institute for Transuranium Elements, a laser-heating and fast radiance spectro-pyrometry facility is used for the laboratory simulation, on a small scale, of NPP core meltdown, the most common type of severe accident (SA) that can occur in a nuclear reactor as a consequence of a failure of the cooling system. This simulation tool permits fast and effective high-temperature measurements on real nuclear materials, such as plutonium and minor actinide-containing fission fuel samples. In this respect, and in its capability to produce large amount of data concerning materials under extreme conditions, the current experimental approach is certainly unique. For current and future concepts of NPP, example results are presented on the melting behavior of some different types of nuclear fuels: uranium-plutonium oxides, carbides, and nitrides. Results on the high-temperature interaction of oxide fuels with containment materials are also briefly shown.

Experimental study of the combined utilization of nuclear power heating plants for big towns and industrial complexes

International Nuclear Information System (INIS)

Neumann, J.; Barabas, K.

1977-01-01

The paper describes a comparison of nuclear power heating plants with an output corresponding to 1000MW(e) with plants of the same output using coal or oil. The economic aspects are compared, both as regards investment and operation costs. The comparison of the environmental aspects is performed on the atmospheric pollution from exhausts and gaseous emission and on the thermal pollutions in hydrosphere and atmosphere. Basic nuclear power plant schemes with two PWRs, each of 1500MW(th), are described. The plant supplies electric power and heat for factories and municipal heating systems (apartments, shops, and other auxiliary municipal facilities). At the same time the basic heat-flow diagram of a nuclear power heating plant is given, together with the relative losses. The study emphasizes the possible utilization of waste heat for heating glasshouses of 200m 2 . The problems of utilizing waste heat, and the needs of a big town and of industrial complexes in the vicinity of the nuclear power heating plant are also considered. (author)

The nuclear winter

International Nuclear Information System (INIS)

Velikhow, Y.P.

1986-01-01

Nuclear winter is an example of possible secondary effects, and if we speak of secondary we are thinking of small-scale second-order effects, but a nuclear winter is not a second-order effect. If you calculate the amount of heat produced by a nuclear explosion, it is a very small amount which does not have any chance of changing the Earth's climate, but a nuclear explosion drives or stars some new mechanism - the mechanism of nuclear winter - after 100 megatons of dust are transferred to the upper atmosphere. Another example of such amplification is radioactive fall-out, especially long-life radioactive fall-out after the possible elimination of the nuclear power industry, nuclear storage and distribution of storage waste around the globe. This is a very powerful amplification mechanism

Aspects of safety and of functional construction and configuration in planning and designing nuclear heating stations

International Nuclear Information System (INIS)

Adam, E.; Mueller, R.; Boettger, M.; Kremtz, U.

1982-01-01

The present studies are based on the design of a technological project of a nuclear heating station with a unit power of 250 MW. Essentially, this nuclear heating station is a three-circuit plant, the primary coolant circuit being based on natural circulation through the reactor vessel with integrated heat exchangers. Starting from the social objective and the derived development structure of the territory, the siting problems in integrating the nuclear heating stations have to be solved. On the basis of the resulting dimensions of the containment the technical and economical specifications of different versions of containment design are evaluated. (author)

On the development of small nuclear power stations

International Nuclear Information System (INIS)

Goetzmann, C.A.

1989-01-01

There are weighty reasons for and against the building of small nuclear power stations. Factors such as specific investment costs, opportunities for and areas of application, geographical conditions as well as those relating to infrastructure, security and availability play an important role in the planning, construction and running of a nuclear power station. For the usual large power stations, the comparatively low specific investment costs and a proven technology are favorable factors which minimize the investment risk. The article presents an overview of reasons for using small power stations and also considers the difficulties which would arise in practice. (orig.) [de

Offshore heat dissipation for nuclear energy centers

International Nuclear Information System (INIS)

Bauman, H.F.

1978-09-01

The technical, environmental, and economic aspects of utilizing the ocean or other large water bodies for the dissipation of reject heat from Nuclear Energy Centers (NECs) were investigated. An NEC in concept is an aggregate of nuclear power plants of 10 GW(e) capacity or greater on a common site. The use of once-through cooling for large power installations offers advantages including higher thermal efficiencies, especially under summer peak-load conditions, compared to closed-cycle cooling systems. A disadvantage of once-through cooling is the potential for greater adverse impacts on the aquatic environment. A concept is presented for minimizing the impacts of such systems by placing water intake and discharge locations relatively distant from shore in deeper water than has heretofore been the practice. This technique would avoid impacts on relatively biologically productive and ecologically sensitive shallow inshore areas. The NEC itself would be set back from the shoreline so that recreational use of the shore area would not be impaired. The characteristics of a heat-dissipation system of the size required for a NEC were predicted from the known characteristics of a smaller system by applying hydraulic scaling laws. The results showed that adequate heat dissipation can be obtained from NEC-sized systems located in water of appropriate depth. Offshore intake and discharge structures would be connected to the NEC pump house on shore via tunnels or buried pipelines. Tunnels have the advantage that shoreline and beach areas would not be disturbed. The cost of an offshore heat-dissipation system depends on the characteristics of the site, particularly the distance to suitably deep water and the type of soil or rock in which water conduits would be constructed. For a favorable site, the cost of an offshore system is estimated to be less than the cost of a closed-cycle system

Floating nuclear heat. And power station 'Pevec' with KLT-40S type reactor plant for remote regions of Russia

International Nuclear Information System (INIS)

Veshnyakov, K.B.; Kiryushin, A.I.; Panov, Yu.K.; Polunichev, V.I.

2000-01-01

Floating small nuclear power plants power for local energy systems of littoral regions of Russia, located far from central energy system, open a new line in nuclear power development. Designing a floating power unit of a lead nuclear heat and power generating station for port Pevec at the Chuckchee national district is currently nearing completion. Most labor-intensive components are being manufactured. The co-generation NPP Pevec is to be created on the basis of a floating power unit with KLT-40S type reactor plant. KLT-40S reactor plant is based on similar propulsion plants, verified at operation of Russia's nuclear-powered civil ships, evolutionary improved by elimination of 'weak points' revealed during its prototypes operation or on the basis of safety analysis. KLT-40S reactor plant uses the most wide-spread and developed in the world practice PWR-type reactor. KLT-40S meets contemporary national and international requirements imposed to future reactor plants. The NHPS description, its main technical-economic data, environmental safety indices, basic characteristics of KLT-40S reactor plant are presented. Prospects of small NPPs utilization outside Russia, particularly as an energy source for sea water desalination, are considered. (author)

Heat supply analysis of steam reforming hydrogen production process in conventional and nuclear

International Nuclear Information System (INIS)

Siti Alimah; Djati Hoesen Salimy

2015-01-01

Tile analysis of heat energy supply in the production of hydrogen by natural gas steam reforming process has been done. The aim of the study is to compare the energy supply system of conventional and nuclear heat. Methodology used in this study is an assessment of literature and analysis based on the comparisons. The study shows that the heat sources of fossil fuels (natural gas) is able to provide optimum operating conditions of temperature and pressure of 850-900 °C and 2-3 MPa, as well as the heat transfer is dominated by radiation heat transfer, so that the heat flux that can be achieved on the catalyst tube relatively high (50-80 kW/m"2) and provide high thermal efficiency of about 85 %. While in the system with nuclear energy, due to the demands of safety, process operating at less than optimum conditions of temperature and pressure of 800-850 °C and 4.5 MPa, as well as the heat transfer is dominated by convection heat transfer, so that the heat flux that can be achieved catalyst tube is relatively low (1020 kW/m"2) and it provides a low thermal efficiency of about 50 %. Modifications of reformer and heat utilization can increase the heat flux up to 40 kW/m"2 so that the thermal efficiency can reach 78 %. Nevertheless, the application of nuclear energy to hydrogen production with steam reforming process is able to reduce the burning of fossil fuels which has implications for the potential decrease in the rate of CO2 emissions into the environment. (author)

Nuclear heating solutions. Realizations and projects

International Nuclear Information System (INIS)

Dumitrescu, Monica; Prisecaru, Ilie

2009-01-01

Considering the present situation of thermal energy in Romania and having in view the fact that Romania is a Kyoto protocol signatory state one estimates that the development of the nuclear energy will have a promising growth. According with the statement of the National Energetic Observer, Romania became a net energy resource importer for the past 30 years and the estimations about the future are not optimistic. The finite reserves of fossil fuel (coal and natural gas), the gradual reduction of their share in the national energy balance with a tendency to become insignificant after 2025, as well as the present situation of the thermal power plants which are already beyond their operation life, all these indicate the nuclear energy as being the most reliable and sustainable future source for thermal energy production. Having in view these circumstances the paper aims at a short presentation of the existing nuclear solutions for district heating. Also, reviewed are the reactor projects that are under different development stage in the world, as well as the best nuclear solutions to be possibly implemented in Romania. The article represents a synthesis of the documentation made by PhD student Monica Dumitrescu in her preparation stage. (authors)

Consideration on nuclear fusion in plasma by the magnetic confinement as a heat engine

International Nuclear Information System (INIS)

Tsuji, Yoshio

1990-01-01

In comparing nuclear fusion in plasma by the magnetic confinement with nuclear fission and chemical reactions, the power density and the function of a heat engine are discussed using a new parameter G introduced as an eigenvalue of a reaction and the value of q introduced to estimate the thermal efficiency of a heat engine. It is shown that the fusion reactor by the magnetic confinement is very difficult to be a modern heat engine because of the lack of some indispensable functions as a modern heat engine. The value of G and q have the important role in the consideration. (author)

Heat capacity of the neutron star inner crust within an extended nuclear statistical equilibrium model

Science.gov (United States)

Burrello, S.; Gulminelli, F.; Aymard, F.; Colonna, M.; Raduta, Ad. R.

2015-11-01

Background: Superfluidity in the crust is a key ingredient for the cooling properties of proto-neutron stars. Present theoretical calculations employ the quasiparticle mean-field Hartree-Fock-Bogoliubov theory with temperature-dependent occupation numbers for the quasiparticle states. Purpose: Finite temperature stellar matter is characterized by a whole distribution of different nuclear species. We want to assess the importance of this distribution on the calculation of heat capacity in the inner crust. Method: Following a recent work, the Wigner-Seitz cell is mapped into a model with cluster degrees of freedom. The finite temperature distribution is then given by a statistical collection of Wigner-Seitz cells. We additionally introduce pairing correlations in the local density BCS approximation both in the homogeneous unbound neutron component, and in the interface region between clusters and neutrons. Results: The heat capacity is calculated in the different baryonic density conditions corresponding to the inner crust, and in a temperature range varying from 100 KeV to 2 MeV. We show that accounting for the cluster distribution has a small effect at intermediate densities, but it considerably affects the heat capacity both close to the outer crust and close to the core. We additionally show that it is very important to consider the temperature evolution of the proton fraction for a quantitatively reliable estimation of the heat capacity. Conclusions: We present the first modelization of stellar matter containing at the same time a statistical distribution of clusters at finite temperature, and pairing correlations in the unbound neutron component. The effect of the nuclear distribution on the superfluid properties can be easily added in future calculations of the neutron star cooling curves. A strong influence of resonance population on the heat capacity at high temperature is observed, which deserves to be further studied within more microscopic calculations.

Efficient power generation from large 7500C heat sources. Application to coal-fired and nuclear power station

International Nuclear Information System (INIS)

Tilliette, Z.P.; Pierre, B.

1980-03-01

Considering the future concern about a more efficient, rational use of heat sources, and also about a greater location flexibility of power plants owing to dry cooling possibility, closed gas cycles can offer new solutions for fossil or nuclear energy. An efficient heat conversion into power is obtained by the combination of a main non-intercooled helium cycle with a flexible, superheated, low pressure bottoming steam cycle. Emphasis is placed on the matching of the two cycle; for that, a recuperator by-pass arrangement is used. The operation of the main gas turbocompressor does not depend upon the operation of the small steam cycle. Results are given for a conservative turbine inlet temperature of 750 0 C. Applications are made to a coal-fired power plant and to a gas turbine, gas-cooled nuclear reactor. Overall net plant efficiencies of 39 per cent and 46 per cent respectively are reached. For a cycle top temperature equal to 850 0 C, corresponding net efficiencies would be 42 and 49 per cent

In core instrumentation for online nuclear heating measurements of material testing reactor

International Nuclear Information System (INIS)

Reynard, C.; Andre, J.; Brun, J.; Carette, M.; Janulyte, A.; Merroun, O.; Zerega, Y.; Lyoussi, A.; Bignan, G.; Chauvin, J-P.; Fourmentel, D.; Glayse, W.; Gonnier, C.; Guimbal, P.; Iracane, D.; Villard, J.-F.

2010-01-01

The present work focuses on nuclear heating. This work belongs to a new advanced research program called IN-CORE which means 'Instrumentation for Nuclear radiations and Calorimetry Online in REactor' between the LCP (University of Provence-CNRS) and the CEA (French Atomic Energy Commission) - Jules Horowitz Reactor (JHR) program. This program started in September 2009 and is dedicated to the conception and the design of an innovative mobile experimental device coupling several sensors and ray detectors for on line measurements of relevant physical parameters (photonic heating, neutronic flux ...) and for an accurate parametric mapping of experimental channels in the JHR Core. The work presented below is the first step of this program and concerns a brief state of the art related to measurement methods of nuclear heating phenomena in research reactor in general and MTR in particular. A special care is given to gamma heating measurements. A first part deals with numerical codes and models. The second one presents instrumentation divided into various kinds of sensor such as calorimeter measurements and gamma ionization chamber measurements. Their basic principles, characteristics such as metrological parameters, operating mode, disadvantages/advantages, ... are discussed. (author)

HTTR demonstration test plan for industrial utilization of nuclear heat

International Nuclear Information System (INIS)

Sato, Hiroyuki; Ohashi, Hirofumi; Yan, Xing L.; Kubo, Shinji; Nishihara, Tetsuo; Tachibana, Yukio; Inagaki, Yoshiyuki

2014-09-01

Japan Atomic Energy Agency has been conducting research and development with a central focus on the utilization of High Temperature engineering Test Reactor (HTTR), the first High Temperature Gas-cooled Reactor (HTGR) in Japan, towards the realization of industrial use of nuclear heat. Several studies have made on the integration of the HTTR with thermochemical iodine-sulfur process and steam methane reforming hydrogen production plant (H 2 plant) as well as helium gas turbine power conversion system. In addition, safety standards for coupling a H 2 plant to a nuclear facility has been investigated. Based on the past design information, the present study identified test items to be validated in the HTTR demonstration test to accomplish a formulation of safety requirement and design consideration for coupling a H 2 plant to a nuclear facility as well as confirmation of overall performance of helium gas turbine system. In addition, plant concepts for the heat utilization system to be connected with the HTTR are investigated. (author)

Hygiene problems in building a nuclear power and heat plant near Bratislava

International Nuclear Information System (INIS)

Chorvat, D.; Mizov, J.; Hladky, E.; Kubik, I.; Carach, J.

1976-01-01

The results are presented of the calculation of the population exposure due to the release of radioactive products from a nuclear power and heating plant accident into the ambient atmosphere (primary coolant circuit rupture) providing the nuclear power and heating plant is sited approximately 500 m from the Slovnaft chemical works in Bratislava. Ground water contamination was analyzed assuming the infiltration of radioactive products from a surface deposit due to fallout and the direct infiltration of the products into the soil in the area of the plant. The results of the assessment of the design basis accident of a WWER-1000 nuclear power and heating plant show unequivocally that the emergency core cooling system, full-pressure containment and the correct function of the containment spray system are able to keep the accident consequences within acceptable limits thus meeting radiation hygiene requirements related to the siting of similar installations in the vicinity of large housing estates. (Oy)

Vitrification of high level nuclear waste inside ambient temperature disposal containers using inductive heating: The SMILE system

International Nuclear Information System (INIS)

Powell, J.; Reich, M.; Barletta, R.

1996-01-01

A new approach, termed SMILE (Small Module Inductively Loaded Energy), for the vitrification of high level nuclear wastes (HLW) is described. Present vitrification systems liquefy the HLW solids and associated frit material in large high temperature melters. The molten mix is then poured into small (∼1 m 3 ) disposal canisters, where it solidifies and cools. SMILE eliminates the separate, large high temperature melter. Instead, the BLW solids and frit melt inside the final disposal containers, using inductive heating. The contents then solidify and cool in place. The SMILE modules and the inductive heating process are designed so that the outer stainless can of the module remains at near ambient temperature during the process cycle. Module dimensions are similar to those of present disposal containers. The can is thermally insulated from the high temperature inner container by a thin layer of refractory alumina firebricks. The inner container is a graphite crucible lined with a dense alumina refractory that holds the HLW and fiit materials. After the SMILE module is loaded with a slurry of HLW and frit solids, an external multi-turn coil is energized with 30-cycle AC current. The enclosing external coil is the primary of a power transformer, with the graphite crucible acting as a single turn ''secondary.'' The induced current in the ''secondary'' heats the graphite, which in turn heats the HLW and frit materials. The first stage of the heating process is carried out at an intermediate temperature to drive off remnant liquid water and water of hydration, which takes about 1 day. The small fill/vent tube to the module is then sealed off and the interior temperature raised to the vitrification range, i.e., ∼1200C. Liquefaction is complete after approximately 1 day. The inductive heating then ceases and the module slowly loses heat to the environment, allowing the molten material to solidify and cool down to ambient temperature

Decontamination and Decommissioning at Small Nuclear Facilities: Facilitating the Submission of Decommissioning Funding Plans

International Nuclear Information System (INIS)

Minor, D.A.; Grumbles, A.

2009-01-01

This paper describes the efforts of the Washington State Department of Health to ensure that small nuclear facilities have the tools each needs to submit Decommissioning Funding Plans. These Plans are required by both the U.S. Nuclear Regulatory Commission (NRC) and in some states - in the case of Washington state, the Washington State Department of Health is the regulator of radioactive materials. Unfortunately, the guidance documents provided by the U.S. NRC pertain to large nuclear facilities, such as nuclear fuel fabrication plants, not the small nuclear laboratory nor small nuclear laundry that may also be required to submit such Plans. These small facilities are required to submit Decommissioning Funding Plans by dint of their nuclear materials inventory, but have only a small staff, such as a Radiation Safety Officer and few authorized users. The Washington State Department of Health and Attenuation Environmental Company have been working on certain tools, such as templates and spreadsheets, that are intended to assist these small nuclear facilities prepare compliant Decommissioning Funding Plans with a minimum of experience and effort. (authors)

Heat exchanger. [Nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Molina, C; Brisseaux, A

1976-01-19

This invention concerns a heat exchanger between a fluid flowing through a tube and a gas. Such an exchanger can be used, inter alia, for removing calories that cannot be used for generating electricity in a thermal or nuclear power station. This exchanger can withstand any pressure surges in the system and even the use of a high vapour pressure coolant such as ammonia, since the fluid flows in a round tube with low pressure drops (both with respect to the fluid to be cooled and the cooling air). It is rigid enough to stand up to being moved and handled as well as to gusts of wind. It is formed of units that can be handled without difficulty and that are easily dismantable and interchangeable, even in service, and it is easily maintained. The exchange area is high for a minimum frontal area and this enables the size of the supporting frame to be reduced and makes it easy to hide it behind a screen of trees should this prove necessary. Finally, it is composed of a small number of standard units thus reducing the industrial production cost. These units are rectangular plates, each one being a flat tubular coil fitted between two flat parallel sheet metal plates having on their outer sides flat top raised bosses. These units are assembled together by the tops of the bosses so as to form an exchanger bank, each bank comprising two collectors to which the bank coils are tightly connected.

Performances of nuclear power plants for combined production of electricity and hot water for district heating

International Nuclear Information System (INIS)

Bronzen, S.

The possibilities for using nuclear power plants for combined production of heat and power seem to be very good in the future. With the chosen 600 MWsub (e) BWR plant a heat output up to 1200 MW can be arranged. An alternative, consisting of steam extractions from the low-pressure turbine, offers a flexible solution for heat and power generation. With this alternative the combined plant can use components from normal condensing nuclear power plants. The flexible extraction design also offers a real possibility for using the combined plant in electric peak generation. However, urban siting requires long distance heat transmission and the pipe design for this transmission is a major problem when planning and optimizing the whole nuclear combined heat and power plant. (author)

Economic and strategic assessment of using nuclear power plants for the district heating. An investigation on the size role

International Nuclear Information System (INIS)

Locatelli, Giorgio; Mancini, Mauro

2011-01-01

Small Medium Reactors (SMR) can play an important role in the global nuclear renaissance coupling the production of Electrical Energy (EE) with by-products useful to increase their economic attractiveness and enhance the public acceptability. Light Water Reactors (LWR) have an average thermal efficiency of about 33%-35%, therefore two third of the thermal energy produced by the nuclear reaction is most of the time wasted. However this heat can be used to warm the buildings in a town representing a cheap and greenhouse free solution for the heating during the winter time. Moreover if the heat is provided free of charge or at low cost to the consumers it can increase the public acceptability of the nuclear power plants. This paper analyses the economic and strategic implications of using SMR instead of LR (Large Reactor) for the nuclear district heating (DH) in the north of Italy. The paper at first assesses the technical requirements of a nuclear DH then its economics versus other traditional forms of warming. Assuming the nuclear options as viable solution a first sizing of the system is proposed, then four cases are presented: the first considers a LR providing 350 MWt, the second a LR providing 1400 MWt, the third 4 SMR providing 350 MWt and the fourth 4 SMR providing 1400 MWt. The four scenarios have been evaluated integrating INCAS (Integrated model for the Competitiveness Assessment of SMRs) with the specific cost elements of the DH. The results point out as the modularity of SMR provides a great advantage in terms of flexibility and backup capacity since each SMR can act as a back-up units respect to the other reactors in case of planned or unplanned outage. Moreover the cost of set-up the DH infrastructure can range from 1% to 9% of the total capital cost of the nuclear plants depending on the size (in terms of power delivered and length of the branches) and the backup capacity. Therefore even if the details of the results are specific for the North Italian

The heat exchanger of small pellet boiler for phytomass

Science.gov (United States)

Mičieta, Jozef; Lenhard, Richard; Jandačka, Jozef

2014-08-01

Combustion of pellets from plant biomass (phytomass) causes various troubles. Main problem is slagging ash because of low melting temperature of ash from phytomass. This problem is possible to solve either improving energetic properties of phytomass by additives or modification of boiler construction. A small-scale boiler for phytomass is different in construction of heat exchanger and furnace mainly. We solve major problem - slagging ash, by decreasing combustion temperature via redesign of pellet burner and boiler body. Consequence of lower combustion temperature is also lower temperature gradient of combustion gas. It means that is necessary to design larger heat exchanging surface. We plane to use underfed burner, so we would utilize circle symmetry heat exchanger. Paper deals design of heat exchanger construction with help of CFD simulation. Our purpose is to keep uniform water flux and combustion gas flux in heat exchanger without zone of local overheating and excess cooling.

Small Punch Tests applied to the nuclear industry

International Nuclear Information System (INIS)

Moreno, M.; Gongora, P.; Bertolino, G; Yawny, A.

2012-01-01

The interest on miniaturized specimen techniques for the characterization of the mechanical behavior of materials was strongly motivated in the early eighties by the different programs associated with the development of fusion reactor technology. The importance of such developments is obvious in the case of the nuclear industry where the limited space available, the presence of fluence gradients in large specimens, the concern about gamma heating and dose to personnel in post-irradiation testing have all been motivations for reducing specimen size. Testing of miniature specimen includes a wide spectrum of techniques such as tensile, instrumented micro-hardness, small punch, bend, fracture, impact and fatigue. Small Punch Testing (SPT) techniques use a spherical penetrator which deforms to failure a miniature disc shaped flat specimen (typically, 3-10 mm in diameter and 0.25-0.50 mm in thickness) supported on its outer rim. Analysis of load-displacement data recorded along the test is performed for the determination of the property of interest. The present work focuses on the characterization of the elastoplastic response of pure Al, ADN 420 structural steel and AISI 304L using SPT and its correlation with the associated standard uniaxial testing behavior. In addition, the sensitivity of the technique to the specific material under study and to different experimental parameters, i.e. specimen diameter and thickness, clearance or clamping force and friction between disc and dies have been assessed both experimentally and by performing simulations using the finite element method (author)

Economic evaluation of heat extraction from nuclear power plants - a criterion for deciding their building order

International Nuclear Information System (INIS)

Navratil, J.

1987-01-01

Heat extraction from nuclear power plants is an important element in the current concept of supplying the population and industries with heat. Economic evaluation of the extraction is one of the factors of the total economic assessment of potential sites for nuclear power plant construction which can contribute to decision making on the priorities of construction. The methodological approach to the assessment of economic contribution of heat extraction from 2x1000 MW nuclear power plant is exemplified using three such sites on the Czechoslovak territory, viz., Opatovice (eastern Bohemia), Blahutovice (northern Moravia), and Kecerovce (eastern Slovakia). The so-called annual converted cost was used as a suitable quantity completely reflecting all significant economic effects of heat extraction. It is shown that the fuel component of the power plant costs is the decisive factor for the amount of the annual converted cost in respect to heat supply and thus also the economic priority of the construction sites of nuclear power plants. (Z.M.). 3 tabs., 3 refs

Possible generation of heat from nuclear fusion in Earth's inner core.

Science.gov (United States)

Fukuhara, Mikio

2016-11-23

The cause and source of the heat released from Earth's interior have not yet been determined. Some research groups have proposed that the heat is supplied by radioactive decay or by a nuclear georeactor. Here we postulate that the generation of heat is the result of three-body nuclear fusion of deuterons confined in hexagonal FeDx core-centre crystals; the reaction rate is enhanced by the combined attraction effects of high-pressure (~364 GPa) and high-temperature (~5700 K) and by the physical catalysis of neutral pions: 2 D +  2 D +  2 D → 2 1 H +  4 He + 2  + 20.85 MeV. The possible heat generation rate can be calculated as 8.12 × 10 12  J/m 3 , based on the assumption that Earth's primitive heat supply has already been exhausted. The H and He atoms produced and the anti-neutrino are incorporated as Fe-H based alloys in the H-rich portion of inner core, are released from Earth's interior to the universe, and pass through Earth, respectively.

Nuclear combined cycle gas turbines for variable electricity and heat using firebrick heat storage and low-carbon fuels

International Nuclear Information System (INIS)

Forsberg, Charles; Peterson, Per F.; McDaniel, Patrick; Bindra, Hitesh

2017-01-01

The world is transitioning to a low-carbon energy system. Variable electricity and industrial energy demands have been met with storable fossil fuels. The low-carbon energy sources (nuclear, wind and solar) are characterized by high-capital-costs and low-operating costs. High utilization is required to produce economic energy. Wind and solar are non-dispatchable; but, nuclear is the dispatchable energy source. Advanced combined cycle gas turbines with firebrick heat storage coupled to high-temperature reactors may enable economic variable electricity and heat production with constant full-power reactor output. Such systems efficiently couple to fluoride-salt-cooled high-temperature reactors (FHRs) with solid fuel and clean salt coolants, molten salt reactors (MSRs) with fuel dissolved in the salt coolant and salt-cooled fusion machines. Open Brayton combined cycles allow the use of natural gas, hydrogen, other fuels and firebrick heat storage for peak electricity production with incremental heat-to-electricity efficiencies from 66 to 70+% efficient. There are closed Brayton cycle options that use firebrick heat storage but these have not been investigated in any detail. Many of these cycles couple to high-temperature gas-cooled reactors (HTGRs). (author)

Heat transfer burnout in tube-type fuel elements of nuclear power reactors

International Nuclear Information System (INIS)

Subbotin, V.; Alexeev, G.; Peskov, O.; Sapankevic, A.

1976-01-01

The conditions are formulated under which the results of the experimental research of the boilino. water heat transfer burnout carried out on models may be applied to fuel elements of nuclear reactors. Experimental material providing data on the heat transfer burnout was expanded by the results of measurements of the uneven (cosine) longitudinal distribution of heat sources. The results of the effects of helical fins or wires on heat transfer burnout are presented. (F.M.)

Heat transfer burnout in tube-type fuel elements of nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Subbotin, V; Alexeev, G; Peskov, O; Sapankevic, A

1976-08-01

The conditions are formulated under which the results of the experimental research of the boiling. water heat transfer burnout carried out on models may be applied to fuel elements of nuclear reactors. Experimental material providing data on the heat transfer burnout was expanded by the results of measurements of the uneven (cosine) longitudinal distribution of heat sources. The results of the effects of helical fins or wires on heat transfer burnout are presented.

Nuclear power and heating plant control rooms. I

International Nuclear Information System (INIS)

Malaniuk, B.

1983-01-01

The questions are discussed of memory capacity, vigilance, speed of data processing, decision-making quality and other demands placed on operators of nuclear power and heating plants. On the example of the accident at the Three Mile Island-2 nuclear power plant, the instants are shown when failure of the human factor owing to a stress situation resulted in the accident not being coped with in time. It is therefore necessary to place high demands on the choice of operators and to devote equal attention to the human factor as to the safety of the technical equipment of the power plant. (J.B.)

After heat removing system of a nuclear reactor

International Nuclear Information System (INIS)

Hayashi, Takao; Yamada, Masao; Ohashi, Kazutaka.

1994-01-01

In a variable conductance heat pipe of an after heat removing system, an evaporation portion and a condensator are connected by a steam diffusing path for an operation fluid and a liquid condensate recycling path. Further, incondensible gases are sealed at the inside together with the operation fluid, and a gas reservoir for the incondensible gases is disposed at the downstream of a condensation portion. If heat input is applied to the evaporation portion of the heat pipe, the incondensible gases are separated to form a boundary between both of them. When the amount of heat applied is small, the incondensible gases partially seal the condensation portion to form a local condensation insensitive portion, so that a heat conductance can be suppressed low. On the other hand, as the amount of heat inputted is increased, the incondensible gases are compressed, the heat conduction area of the condensation portion is increased and a heat conductance is increased to conduct self-control so as to increase heat transfer performance of the heat pipe. Then, the liquid condensate is recycled to the evaporation portion by spontaneous dripping of the condensate itself without wick, thereby enabling to conduct automatic switching so as to increase the heat dissipation amount to maximum. (N.H.)

Examining the intersection between splicing, nuclear export and small RNA pathways.

Science.gov (United States)

Nabih, Amena; Sobotka, Julia A; Wu, Monica Z; Wedeles, Christopher J; Claycomb, Julie M

2017-11-01

Nuclear Argonaute/small RNA pathways in a variety of eukaryotic species are generally known to regulate gene expression via chromatin modulation and transcription attenuation in a process known as transcriptional gene silencing (TGS). However, recent data, including genetic screens, phylogenetic profiling, and molecular mechanistic studies, also point to a novel and emerging intersection between the splicing and nuclear export machinery with nuclear Argonaute/small RNA pathways in many organisms. In this review, we summarize the field's current understanding regarding the relationship between splicing, export and small RNA pathways, and consider the biological implications for coordinated regulation of transcripts by these pathways. We also address the importance and available approaches for understanding the RNA regulatory logic generated by the intersection of these particular pathways in the context of synthetic biology. The interactions between various eukaryotic RNA regulatory pathways, particularly splicing, nuclear export and small RNA pathways provide a type of combinatorial code that informs the identity ("self" versus "non-self") and dictates the fate of each transcript in a cell. Although the molecular mechanisms for how splicing and nuclear export impact small RNA pathways are not entirely clear at this early stage, the links between these pathways are widespread across eukaryotic phyla. The link between splicing, nuclear export, and small RNA pathways is emerging and establishes a new frontier for understanding the combinatorial logic of gene regulation across species that could someday be harnessed for therapeutic, biotechnology and agricultural applications. This article is part of a Special Issue entitled "Biochemistry of Synthetic Biology - Recent Developments" Guest Editor: Dr. Ilka Heinemann and Dr. Patrick O'Donoghue. Copyright © 2017 Elsevier B.V. All rights reserved.

Initial crystallographic studies of a small heat-shock protein from Xylella fastidiosa

International Nuclear Information System (INIS)

Tada, Susely F. S.; Saraiva, Antonio Marcos; Lorite, Gabriela S.; Rosselli-Murai, Luciana K.; Pelloso, Alexandre César; Santos, Marcelo Leite dos; Trivella, Daniela B. B.; Cotta, Mônica A.; Souza, Anete Pereira de; Aparicio, Ricardo

2012-01-01

Initial crystallographic studies of the X. fastidiosa small heat-shock protein HSP17.9 are reported. The ORF XF2234 in the Xylella fastidiosa genome was identified as encoding a small heat-shock protein of 17.9 kDa (HSP17.9). HSP17.9 was found as one of the proteins that are induced during X. fastidiosa proliferation and infection in citrus culture. Recombinant HSP17.9 was crystallized and surface atomic force microscopy experiments were conducted with the aim of better characterizing the HSP17.9 crystals. X-ray diffraction data were collected at 2.7 Å resolution. The crystal belonged to space group P4 3 22, with unit-cell parameters a = 68.90, b = 68.90, c = 72.51 Å, and is the first small heat-shock protein to crystallize in this space group

Analysis on Domestic Law and Management Trend Related to Small-Quantity Nuclear Material

International Nuclear Information System (INIS)

Park, Jae Beom; Lee, Kyong Woo; Shim, Hye Won; Min, Gyung Sik

2005-01-01

International Atomic Energy Agency (IAEA) has requested Korea to establish and manage the law ruling all nuclear materials through the INFCIRC/153. Now, it has been 30 years since Korea made the agreement, INFCIRC/153, with IAEA. Korea has tried their best to accomplish the international standard in nuclear control field and it is a fact that Korea finally produced some results in the nuclear control field. Related to nuclear material control, Korea is above the common level appropriately ranked 6th in the world in terms of nuclear power. Before 2000, Korea was making the foundation secure in the nuclear control. IAEA did not urge to establish the law supervising the small-quantity nuclear material and depleted uranium (DU). In a turnaround from early IAEA moderate line to Korea, the situation was changed. Since IAEA brought up the agenda to 2000 Joint Review Meeting between Korea-IAEA, IAEA has asked Korea to establish the control system for smallquantity nuclear material and DU. In 2003, the Korean government set up a project establishing the control system about all nuclear material including small-quantity nuclear material and DU. National Nuclear Management and Control Agency (NNCA), delegating the business relating to international controlling materials from government, developed some modules in nuclear material control system and operated it. The system includes a controlling system for small-quantity nuclear material. NNCA on behalf of government has collected the information and Korea Ministry of Science and Technology (MOST) has reported the information to the IAEA. This paper introduces you the background of controlling the small-quantity nuclear material and the system of controlling nuclear material in Korea. And it will suggest the improvement of the management method in the system for small-quantity nuclear material

Analysis on Domestic Law and Management Trend Related to Small-Quantity Nuclear Material

Energy Technology Data Exchange (ETDEWEB)

Park, Jae Beom; Lee, Kyong Woo; Shim, Hye Won; Min, Gyung Sik [National Nuclear Management and Control Agency, Daejeon (Korea, Republic of)

2005-07-01

International Atomic Energy Agency (IAEA) has requested Korea to establish and manage the law ruling all nuclear materials through the INFCIRC/153. Now, it has been 30 years since Korea made the agreement, INFCIRC/153, with IAEA. Korea has tried their best to accomplish the international standard in nuclear control field and it is a fact that Korea finally produced some results in the nuclear control field. Related to nuclear material control, Korea is above the common level appropriately ranked 6th in the world in terms of nuclear power. Before 2000, Korea was making the foundation secure in the nuclear control. IAEA did not urge to establish the law supervising the small-quantity nuclear material and depleted uranium (DU). In a turnaround from early IAEA moderate line to Korea, the situation was changed. Since IAEA brought up the agenda to 2000 Joint Review Meeting between Korea-IAEA, IAEA has asked Korea to establish the control system for smallquantity nuclear material and DU. In 2003, the Korean government set up a project establishing the control system about all nuclear material including small-quantity nuclear material and DU. National Nuclear Management and Control Agency (NNCA), delegating the business relating to international controlling materials from government, developed some modules in nuclear material control system and operated it. The system includes a controlling system for small-quantity nuclear material. NNCA on behalf of government has collected the information and Korea Ministry of Science and Technology (MOST) has reported the information to the IAEA. This paper introduces you the background of controlling the small-quantity nuclear material and the system of controlling nuclear material in Korea. And it will suggest the improvement of the management method in the system for small-quantity nuclear material.

Process for the transport of heat energy released by a nuclear reactor

International Nuclear Information System (INIS)

Nuernberg, H.W.; Wolff, G.

1978-01-01

The heat produced in a nuclear reactor is converted into latent chemical binding energy. The heat can be released again below 400 0 C by recombination after transport by decomposition of ethane or propane into ethylene or propylene and hydrogen. (TK) [de

Possible uses of nuclear energy in central heating of Ankara

International Nuclear Information System (INIS)

Agirsoy, L.

1987-01-01

In this master thesis, a study was carried out for the district heating of the plateau region where the population and air pollution densities are the highest. First the heat requirements of differently populated regions were calculated, then by taking different temperature decreases of hot water in buildings; flow rates, pipe diameters and pressure losses corres-ponding to these temperature decreases were obtained. An optimum division of total heat load as peak and base loads was studied and it was seen that the unit heat cost could be lowered by employing two stations for the heating of buildings. The optimum division and unit heat cost calculations were carried out for various alternative heating systems and it was seen that nuclear combined cycle base-load station and a peak-load station operating on fuel-oil was obtained to be the most advantageous system from an economic point of view. (author)

Mathematical modelling of heat production in deep geological repository of high-level nuclear waste

International Nuclear Information System (INIS)

Kovanda, O.

2017-01-01

Waste produced by nuclear industry requires special handling. Currently, there is a research taking place, focused at possibilities of nuclear waste storage in deep geological repositories, hosted in stable geological environment. The high-level nuclear waste produces significant amount of heat for a long time, which can affect either environment outside of or within the repository in a negative way. Therefore to reduce risks, it is desirable to know the principles of such heat production, which can be achieved using mathematical modeling. This thesis comes up with a general model of heat production-time dependency, dependable on initial composition of the waste. To be able to model real situations, output of this thesis needs to be utilized in an IT solution. (authors)

Development program for the high-temperature nuclear process heat system

International Nuclear Information System (INIS)

Jiacoletti, R.J.

1975-09-01

A comprehensive development program plan for a high-temperature nuclear process heat system with a very high temperature gas-cooled reactor heat source is presented. The system would provide an interim substitute for fossil-fired sources and ultimately the vehicle for the production of substitute and synthetic fuels to replace petroleum and natural gas. The dwindling domestic reserves of petroleum and natural gas dictate major increases in the utilization of coal and nuclear sources to meet the national energy demand. The nuclear process heat system has significant potential in a unique combination of the two sources that is environmentally and economically attractive and technically sound: the production of synthetic fuels from coal. In the longer term, it could be the key component in hydrogen production from water processes that offer a substitute fuel and chemical feedstock free of dependence on fossil-fuel reserves. The proposed development program is threefold: a process studies program, a demonstration plant program, and a supportive research and development program. Optional development scenarios are presented and evaluated, and a selection is proposed and qualified. The interdependence of the three major program elements is examined, but particular emphasis is placed on the supportive research and development activities. A detailed description of proposed activities in the supportive research and development program with tentative costs and schedules is presented as an appendix with an assessment of current status and planning

Experimental investigation on Heat Transfer Performance of Annular Flow Path Heat Pipe

International Nuclear Information System (INIS)

Kim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin; Bang, In Cheol

2015-01-01

Mochizuki et al. was suggested the passive cooling system to spent nuclear fuel pool. Detail analysis of various heat pipe design cases was studied to determine the heat pipes cooling performance. Wang et al. suggested the concept PRHRS of MSR using sodium heat pipes, and the transient performance of high temperature sodium heat pipe was numerically simulated in the case of MSR accident. The meltdown at the Fukushima Daiichi nuclear power plants alarmed to the dangers of station blackout (SBO) accident. After the SBO accident, passive decay heat removal systems have been investigated to prevent the severe accidents. Mochizuki et al. suggested the heat pipes cooling system using loop heat pipes for decay heat removal cooling and analysis of heat pipe thermal resistance for boiling water reactor (BWR). The decay heat removal systems for pressurized water reactor (PWR) were suggested using natural convection mechanisms and modification of PWR design. Our group suggested the concept of a hybrid heat pipe with control rod as Passive IN-core Cooling System (PINCs) for decay heat removal for advanced nuclear power plant. Hybrid heat pipe is the combination of the heat pipe and control rod. In the present research, the main objective is to investigate the effect of the inner structure to the heat transfer performance of heat pipe containing neutron absorber material, B 4 C. The main objective is to investigate the effect of the inner structure in heat pipe to the heat transfer performance with annular flow path. ABS pellet was used instead of B 4 C pellet as cylindrical structures. The thermal performances of each heat pipes were measured experimentally. Among them, concentric heat pipe showed the best performance compared with others. 1. Annular evaporation section heat pipe and annular flow path heat pipe showed heat transfer degradation. 2. AHP also had annular vapor space and contact cooling surface per unit volume of vapor was increased. Heat transfer coefficient of

Glas generator for the steam gasification of coal with nuclear generated heat

International Nuclear Information System (INIS)

Buchner, G.

1980-01-01

The use of heat from a High Temperature Reactor (HTR) for the steam gasification of coal saves coal, which otherwise is burnt to generate the necessary reaction heat. The gas generator for this process, a horizontal pressure vessel, contains a fluidized bed of coal and steam. An ''immersion-heater'' type of heat exchanger introduces the nuclear generated heat to the process. Some special design problems of this gasifier are presented. Reference is made to the present state of development of the steam gasification process with heat from high temperature reactors. (author)

Design of SES-10 nuclear reactor for district heating

International Nuclear Information System (INIS)

Cuttler, J.M.

1991-03-01

The SES-10 units are unpressurized, pool-type nuclear reactors of 10MW rating, designed for supplying energy to hot water district heating systems, economically and without pollution. Water for heat distribution is brought to a maximum temperature of 85 degrees C. Conventional heating units supplement the output from SES-10 units for peak load and during maintenance. The SES-10 is housed in a low-cost building, with a double-walled pool in the ground. A naturally circulating primary system and a pumped secondary system transport heat from the reactor to the distribution system. The unit is fully automated and easy to maintain. Because of the many active and passive safety features, it is feasible to license the SES-10 for operation in a city and easy to explain it to the public for their acceptance. The core lasts approximately 43 months at a capacity factor of 70%, and the cost of heat is expected to be 2 to 2.5 cents/kWh

A Single Global Small-User Nuclear Repository

International Nuclear Information System (INIS)

Conca, J.L.; Wright, J.

2009-01-01

Global energy partnerships in nuclear power, proposed by France, Russia, U.S. and England, seek to address the proliferation issue by controlling fuel production and nuclear materials, removing the need for each country to develop enrichment, fabrication, recycling or disposal capabilities. Several of the large generator countries such as France, the U.S., Japan, S. Korea, Russia, the U.K., China and India, all have plans for deep geologic repositories because they anticipate sufficient waste over the next century to justify the expense of a repository. However, countries having, or planning, less than five reactors, such as Egypt, Iran, Indonesia, Brazil and about 30 other countries, will not have sufficient waste generation, or a favorable geologic site, to justify the economic and environmental issues of developing their own repository. The Salado salt formation in New Mexico, set aside for nuclear waste disposal within the 16 square-mile area by the Land Withdrawal Act of 1992, is the most optimal geologic formation for the permanent disposal of any nuclear waste and is easily able to host all of the commercial nuclear waste that will be generated in the next thousand years. The U.S. commercial nuclear waste needs presently surpass all others, and will for the foreseeable future. Hosting the relatively small amount of waste from these small-user nations will add little to U.S. waste stream while the cost/benefit analysis from the standpoint of operations, safety, geology, cost and proliferation is overwhelmingly positive for developing such a global repository. Oceanic and overland transportation, high-level disposal logistics and costs from several programs, including WIPP, have demonstrated that the operation would pay for itself from international user fees with no U.S. taxpayer dollars required and still save the world about $400 billion over 100 years. The ethical considerations alone are compelling. (authors)

High-Efficiency Small-Scale Combined Heat and Power Organic Binary Rankine Cycles

OpenAIRE

Costante Mario Invernizzi; Nadeem Ahmed Sheikh

2018-01-01

Small-CHP (Combined Heat and Power) systems are generally considered a valuable technological option to the conventional boilers, in a technology developed context. If small-CHP systems are associated with the use of renewable energies (biomass, for example) they could play an important role in distributed generation even in developing countries or, in any case, where there are no extensive electricity networks. Traditionally the considered heat engines for micro- or mini-CHP are: the gas eng...

Europairs project: creating an alliance of nuclear and non-nuclear industries for developing nuclear cogeneration

International Nuclear Information System (INIS)

Hittner, Dominique; Bogusch, Edgar; Viala, Celine; Angulo, Carmen; Chauvet, Vincent; Fuetterer, Michael A.; De Groot, Sander; Von Lensa, Werner; Ruer, Jacques; Griffay, Gerard; Baaten, Anton

2010-01-01

Developers of High Temperature Reactors (HTR) worldwide acknowledge that the main asset for market breakthrough is its unique ability to address growing needs for industrial cogeneration of heat and power (CHP) owing to its high operating temperature and flexibility, adapted power level, modularity and robust safety features. HTR are thus well suited to most of the non-electric applications of nuclear energy, which represent about 80% of total energy consumption. This opens opportunities for reducing CO 2 emissions and securing energy supply which are complementary to those provided by systems dedicated to electricity generation. A strong alliance between nuclear and process heat user industries is a necessity for developing a nuclear system for the conventional process heat market, much in the same way as the electronuclear development required a close partnership with utilities. Initiating such an alliance is one of the objectives of the EUROPAIRS project just started in the frame of the EURATOM 7. Framework Programme (FP7) under AREVA coordination. Within EUROPAIRS, process heat user industries express their requirements whereas nuclear industry will provide the performance window of HTR. Starting from this shared information, an alliance will be forged by assessing the feasibility and impact of nuclear CHP from technical, industrial, economical, licensing and sustainability perspectives. This assessment work will allow pointing out the main issues and challenges for coupling an HTR with industrial process heat applications. On this basis, a Road-map will be elaborated for achieving an industrially relevant demonstration of such a coupling. This Road-map will not only take into consideration the necessary nuclear developments, but also the required adaptations of industrial application processes and the possible development of heat transport technologies from the nuclear heat source to application processes. Although only a small and short project (21 months

Nuclear heat-load limits for above-grade storage of solid transuranium wastes

International Nuclear Information System (INIS)

Clontz, B.G.

1978-06-01

Nuclear safety and heat load limits were established for above-grade storage of transuranium (TRU) wastes. Nuclear safety limits were obtained from a study by J.L. Forstner and are summarized. Heat load limits are based on temperature calculations for TRU waste drums stored in concrete containers (hats), and results are summarized. Waste already in storage is within these limits. The limiting factors for individual drum heat load limits were (1) avoidance of temperatures in excess of 190 0 F (decomposition temperature of anion resin) when anion resin is present in a concrete hat, and (2) avoidance of temperatures in excess of 450 0 F (ignition temperature of paper) at any point inside a waste drum. The limiting factor for concrete had heat load limits was avoidance of temperatures in excess of 265 0 F (melt point of high density polyethylene) at the drum liners. A temperature profile for drums and hats filled to recommended limits is shown. Equations and assumptions used were conservative

Production of synthesis gas and methane via coal gasification utilizing nuclear heat

International Nuclear Information System (INIS)

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

1982-01-01

The steam gasificaton of coal requires a large amount of energy for endothermic gasification, as well as for production and heating of the steam and for electricity generation. In hydrogasification processes, heat is required primarily for the production of hydrogen and for preheating the reactants. Current developments in nuclear energy enable a gas cooled high temperature nuclear reactor (HTR) to be the energy source, the heat produced being withdrawn from the system by means of a helium loop. There is a prospect of converting coal, in optimal yield, into a commercial gas by employing the process heat from a gas-cooled HTR. The advantages of this process are: (1) conservation of coal reserves via more efficient gas production; (2) because of this coal conservation, there are lower emissions, especially of CO 2 , but also of dust, SO 2 , NO/sub x/, and other harmful substances; (3) process engineering advantages, such as omission of an oxygen plant and reduction in the number of gas scrubbers; (4) lower gas manufacturing costs compared to conventional processes. The main problems involved in using nuclear energy for the industrial gasification of coal are: (1) development of HTRs with helium outlet temperatures of at least 950 0 C; (2) heat transfer from the core of the reactor to the gas generator, methane reforming oven, or heater for the hydrogenation gas; (3) development of a suitable allothermal gas generator for the steam gasification; and (4) development of a helium-heated methane reforming oven and adaption of the hydrogasification process for operation in combination with the reactor. In summary, processes for gasifying coal that employ heat from an HTR have good economic and technical prospects of being realized in the future. However, time will be required for research and development before industrial application can take place. 23 figures, 4 tables. (DP)

ITER Generic Diagnostic Upper Port Plug Nuclear Heating and Personnel Dose Rate Assessment

International Nuclear Information System (INIS)

Feder, Russell E.; Youssef, Mahmoud Z.

2009-01-01

Neutronics analysis to find nuclear heating rates and personnel dose rates were conducted in support of the integration of diagnostics in to the ITER Upper Port Plugs. Simplified shielding models of the Visible-Infrared diagnostic and of a large aperture diagnostic were incorporated in to the ITER global CAD model. Results for these systems are representative of typical designs with maximum shielding and a small aperture (Vis-IR) and minimal shielding with a large aperture. The neutronics discrete-ordinates code ATTILA(reg s ign) and SEVERIAN(reg s ign) (the ATTILA parallel processing version) was used. Material properties and the 500 MW D-T volume source were taken from the ITER 'Brand Model' MCNP benchmark model. A biased quadrature set equivalent to Sn=32 and a scattering degree of Pn=3 were used along with a 46-neutron and 21-gamma FENDL energy subgrouping. Total nuclear heating (neutron plug gamma heating) in the upper port plugs ranged between 380 and 350 kW for the Vis-IR and Large Aperture cases. The Large Aperture model exhibited lower total heating but much higher peak volumetric heating on the upper port plug structure. Personnel dose rates are calculated in a three step process involving a neutron-only transport calculation, the generation of activation volume sources at pre-defined time steps and finally gamma transport analyses are run for selected time steps. ANSI-ANS 6.1.1 1977 Flux-to-Dose conversion factors were used. Dose rates were evaluated for 1 full year of 500 MW DT operation which is comprised of 3000 1800-second pulses. After one year the machine is shut down for maintenance and personnel are permitted to access the diagnostic interspace after 2-weeks if dose rates are below 100 (micro)Sv/hr. Dose rates in the Visible-IR diagnostic model after one day of shutdown were 130 (micro)Sv/hr but fell below the limit to 90 (micro)Sv/hr 2-weeks later. The Large Aperture style shielding model exhibited higher and more persistent dose rates. After 1

Technical management on commissioning test of nuclear heating reactor

International Nuclear Information System (INIS)

Zhang Yajun; Su Qingshan

1999-01-01

The commissioning is the last construction stage of a nuclear heating project. The commissioning quality will directly affect on the safe operation and availability of the heating reactor. The author presents the whole test process until the completion of the test report from the point of test documents, including the preparation and execution of the test, the management of the various unexpected events during the test. And it will be emphatically discussed that the managing procedures of the various unexpected events during the test, including temporary control change, setpoint change, unexpected events and design change

An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

International Nuclear Information System (INIS)

Clark, J.S.; Walton, J.T.; Mcguire, M.L.

1992-07-01

Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines. 11 refs

Nuclear-heat deposition for a fusion-like neutron environment

International Nuclear Information System (INIS)

Carter, L.L.; Hegberg, D.E.; Wilcox, A.D.

1981-10-01

Calculated nuclear heat deposition profiles within the thermal shield of the FMIT facility are sensitive to the cross-section data base - particularly an energy balance consistency between gamma production cross-sections and neutron KERMA factors. Infinite medium calculations were made with the Monte Carlo code to provide integral validations of energy balances relevant to this aspect of the data base. Inconsistencies were found and corrected. There was also concern about the adequacy of the high energy cross sections (10 MeV < E < 30 MeV) for the moderation and transport of the (d,Li) source neutrons. A preliminary analysis of a measurement with a (d,Li) source in the center of an iron block has improved our confidence in the high energy cross section - data base for this application. Monte Carlo calculations have been utilized to calculate three-dimensional profiles of nuclear heat deposition. Representative profiles were displayed for two walls of the FMIT test cell

Environmental issues and competitiveness of district heating systems

International Nuclear Information System (INIS)

Kypreos, S.

1991-01-01

The advantages of district heating systems are evaluated in competition to individual heating for the Swiss markets. The preservation of the environmental quality on the national (clean air concept) and global scale (Toronto recommendation) is formulated as constraint of the energy system. The implications of these constraints for the economic competition of district heating is evaluated. The study estimates the evolution of energy demand in the heating markets and shortly describes the technical possibilities in satisfying demand by a set of conventional heating systems, systems using renewable energy sources, energy conservation measures and district heating systems based on conventional or nuclear energy sources. The main conclusion is that small capacity nuclear district heating systems, if acceptable, could enhance the flexibility of the Swiss energy system in respect to CO 2 control. (author) 3 figs., 4 tabs., 9 refs

Advances in processing technologies for titanium heat exchanger tubes of fossil and nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Likhareva, T P; Tchizhik, A A; Chavchanidze, N N [Polzanov Central Boiler and Turbine Institute, St. Petersburg (Russian Federation)

1999-12-31

The advances in processing technologies for titanium heat exchangers with rolled and welded tubes of fossil and nuclear power plants in Russia are presented. The special methodology of investigations with constant small strain rate have been used to study the effects of mixed corrosion and creep processes in condensers cooled by sea or synthetic sea waters. The results of corrosion creep tests and K1scc calculations are given. The Russian science activities concerning condensers manufactured from titanium show the possibilities for designing structures with very high level service reliability in different corrosion aggressive mediums with high total salt, Cl-ion and oxygen contents. (orig.)

Advances in processing technologies for titanium heat exchanger tubes of fossil and nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Likhareva, T.P.; Tchizhik, A.A.; Chavchanidze, N.N. [Polzanov Central Boiler and Turbine Institute, St. Petersburg (Russian Federation)

1998-12-31

The advances in processing technologies for titanium heat exchangers with rolled and welded tubes of fossil and nuclear power plants in Russia are presented. The special methodology of investigations with constant small strain rate have been used to study the effects of mixed corrosion and creep processes in condensers cooled by sea or synthetic sea waters. The results of corrosion creep tests and K1scc calculations are given. The Russian science activities concerning condensers manufactured from titanium show the possibilities for designing structures with very high level service reliability in different corrosion aggressive mediums with high total salt, Cl-ion and oxygen contents. (orig.)

Ultimate after-heat removal system for nuclear reactors

International Nuclear Information System (INIS)

Bernard, L. Jr.

1980-01-01

The invention concerns the safety region of a nuclear power plant, especially the divertor for the residual heat which keeps forming after shutdown of the reactor. According to the invention a dry cooling tower of enclosed construction is planned. The walls and roof shall be rocket-proof. Such a configuration is described and explained by means of designs. (UWI) [de

Base-Load and Peak Electricity from a Combined Nuclear Heat and Fossil Combined-Cycle Plant

International Nuclear Information System (INIS)

Conklin, Jim; Forsberg, Charles W.

2007-01-01

A combined-cycle power plant is proposed that uses heat from a high-temperature reactor and fossil fuel to meet base-load and peak electrical demands. The high-temperature gas turbine produces shaft power to turn an electric generator. The hot exhaust is then fed to a heat recovery steam generator (HRSG) that provides steam to a steam turbine for added electrical power production. A simplified computational model of the thermal power conversion system was developed in order to parametrically investigate two different steady-state operation conditions: base load nuclear heat only from an Advanced High Temperature Reactor (AHTR), and combined nuclear heat with fossil heat to increase the turbine inlet temperature. These two cases bracket the expected range of power levels, where any intermediate power level can result during electrical load following. The computed results indicate that combined nuclear-fossil systems have the potential to offer both low-cost base-load electricity and lower-cost peak power relative to the existing combination of base-load nuclear plants and separate fossil-fired peak-electricity production units. In addition, electric grid stability, reduced greenhouse gases, and operational flexibility can also result with using the conventional technology presented here for the thermal power conversion system coupled with the AHTR

Evaluation of external heat loss from a small-scale expander used in organic Rankine cycle

International Nuclear Information System (INIS)

Li Jing; Pei Gang; Li Yunzhu; Ji Jie

2011-01-01

With the scaling down of the Organic Rankine Cycle (ORC), the engine shaft power is not only determined by the enthalpy drop in the expansion process but also the external heat loss from the expander. Theoretical and experimental support in evaluating small-scale expander heat loss is rare. This paper presents a quantitative study on the convection, radiation, and conduction heat transfer from a kW-scale expander. A mathematical model is built and validated. The results show that the external radiative or convective heat loss coefficient was about 3.2 or 7.0 W/K.m 2 when the ORC operated around 100 o C. Radiative and convective heat loss coefficients increased as the expander operation temperature increased. Conductive heat loss due to the connection between the expander and the support accounted for a large proportion of the total heat loss. The fitting relationships between heat loss and mean temperature difference were established. It is suggested that low conductivity material be embodied in the support of expander. Mattress insulation for compact expander could be eliminated when the operation temperature is around 100 o C. - Highlights: → A close examination of external heat loss from a small expander is presented. → Theoretical analysis and experimental test were conducted. → The established formulas can be applied to other small ORC expanders. → The results are useful in further research of small-scale ORC.

Potential of small nuclear reactors for future clean and safe energy sources

International Nuclear Information System (INIS)

Sekimoto, H.

1992-01-01

To cope with the various kinds of energy demands expected in the 21st century, it is necessary to explore the potential of small nuclear reactors and to find a way of promoting their introduction to society. The main goal of current research activities is 'the constitution of the self-consistent nuclear energy system'. These activities can be understood by realizing that the nuclear community is facing a turning point for its survival in the 21st century. Self-consistency can be manifested by investigating and developing the potential advantages of the nuclear fission reaction and lessening the potential disadvantages. The contributions in this volume discuss concepts of small reactors, applications of small reactors, and consistency with conventional energy supply systems

Development, calibration and experimental results obtained with an innovative calorimeter (CALMOS) for nuclear heating measurements

International Nuclear Information System (INIS)

Carcreff, H.; Cloute-Cazalaa, V.; Salmon, L.

2011-01-01

Nuclear heating inside an MTR reactor has to be known in order to be able to control samples temperature during irradiation experiments. An R and D program has been carried out at CEA to design a new type of in-core calorimetric system. This new development, started in 2002, has for main objective to manufacture a calorimeter suitable to monitoring nuclear heating inside the 70 MWth OSIRIS material testing reactor operated by CEA's Nuclear Energy Div. at the Saclay research center. An innovative calorimetric probe, associated to a specific handling system, has been designed to provide access to measurements both along the fissile height and on the upper part of the core, where nuclear heating still remains high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for process validation, while a displacement system has been especially studied to move the probe along a given axial measurement range. This paper deals with the development, tests on preliminary mock-ups and the finalization of the probe. Main modeling and experimental results are presented. Moreover, alternative methods to calibration for nuclear heating rate measurements which are now possible with this new calorimeter are presented and discussed. (authors)

Development, calibration, and experimental results obtained with an innovative calorimeter (CALMOS) for nuclear heating measurements

International Nuclear Information System (INIS)

Carcreff, Hubert; Cloute-Cazalaa, Veronique; Salmon, Laurent

2012-01-01

Nuclear heating inside an MTR reactor has to be known in order to be able to control samples temperature during irradiation experiments. An R and D program has been carried out at CEA to design a new type of in-core calorimetric system. This new development, started in 2002, has for main objective to manufacture a calorimeter suitable to monitoring nuclear heating inside the 70 MWth OSIRIS material testing reactor operated by CEA's Nuclear Energy Division at the Saclay research center. An innovative calorimetric probe, associated to a specific handling system, has been designed to provide access to measurements both along the fissile height and on the upper part of the core, where nuclear heating still remains high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for process validation, while a displacement system has been especially studied to move the probe along a given axial measurement range. This paper deals with the development, tests on preliminary mock-ups and the finalization of the probe. Main modeling and experimental results are presented. Moreover, alternative methods to calibration for nuclear heating rate measurements which are now possible with this new calorimeter are presented and discussed. (authors)

Small Modular Reactors for Countries with Small to Medium Electric Grids - An Economically Sensible Solution

International Nuclear Information System (INIS)

Young, P.

2012-01-01

There has recently been a renewed interest throughout the world in small nuclear units for generating electricity and for other applications. A report by the World Nuclear Association discussing the advantages of small modular nuclear reactors (SMRs) over traditional nuclear reactor designs, states that ''modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production, and reduced siting costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction.'' Since the inception of nuclear power, the size of reactor units has grown from under 100 MWe to more than 1600 MWe. Today, due partly to the high capital cost of large power reactors and partly to the need to service small electricity grids, there is a move to develop smaller units. These may be built individually or as modules in a larger plant. SMRs are a good fit in markets where anticipated electricity demand is projected to increase incrementally, because SMRs could be built in series as needed. SMRs might be particularly attractive in countries that currently rely on diesel generators for producing electricity. Small reactors could make economic sense because of the high cost of diesel generation compared to the low marginal cost of producing electricity from nuclear energy. (Keeping in mind the initial investment costs and the need to establish a national regulatory program.) Some SMR designs are fabricated in a factory and then delivered to the site. This could be a solution for markets that lack the qualified engineers and skilled craft workers needed to construct large reactors on site. This paper will provide an overview of the types and attributes of SMRs in use or under development worldwide, describe the similarities and important differences between designs, discuss potential applications for SMRs, including baseload electricity generation, electricity generation for remote locations and areas with

A trend to small nuclear power plants?

International Nuclear Information System (INIS)

Lameira, Fernando Soares

2000-01-01

The release of fossil fuel greenhouse gases and the depletion of cheap oil reserves outside the Persic Gulf suggest a promising scenario for the future of nuclear power. But the end of the Cold War, the crisis of the state, axiological questions and globalization may lead to a marked for small power plants. The purpose of this paper is to analyze these factors, since they are not always considered all together in the future scenarios for nuclear power. It is concluded that the current evolutionary trend of nuclear power projects toward big plants may become one of the main barriers for the introduction of new plants in the future. It is suggested that a combination of fission reactors with technologies unavailable in the 1950's, when the design characteristics of the current nuclear power plants were established, could be considered to overcome this barrier. (author)

Study of passive residual heat removal system of a modular small PWR reactor; Estudo do sistema passivo de remoção de calor residual de um reator PWR pequeno modular

Energy Technology Data Exchange (ETDEWEB)

Araujo, Nathália N., E-mail: nathalianunes@poli.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Departamento de Engenharia Nuclear; Faccini, José L.H., E-mail: faccini@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Su, Jian, E-mail: sujian@lasme.coppe.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

2017-07-01

This paper presents a study on the passive residual heat removal system (PRHRS) of a small modular nuclear reactor (SMR) of 75MW. More advanced nuclear reactors, such as generation III + and IV, have passive safety systems that automatically go into action in order to prevent accidents. The purpose of the PRHRS is to transfer the decay heat from the reactor's nuclear fuel, keeping the core cooled after the plant has shut down. It starts operating in the event of fall of power supply to the nuclear station, or in the event of an unavailability of the steam generator water supply system. Removal of decay heat from the core of the reactor is accomplished by the flow of the primary refrigerant by natural circulation through heat exchangers located in a pool filled with water located above the core. The natural circulation is caused by the density gradient between the reactor core and the pool. A thermal and comparative analysis of the PRHRS was performed consisting of the resolution of the mass conservation equations, amount of movement and energy and using incompressible fluid approximations with the Boussinesq approximation. Calculations were performed with the aid of Mathematica software. A design of the heat exchanger and the cooling water tank was done so that the core of the reactor remained cooled for 72 hours using only the PRHRS.

The nuclear desalination project in Morocco

International Nuclear Information System (INIS)

1996-01-01

The objectives of the seawater desalination demonstration plant in Morocco are to buildup the technical confidence in the utilization of nuclear heating reactor for seawater desalination; to establish a data base for reliable extrapolation of water production costs for a commercial nuclear plant; and to further strengthen the nuclear infrastructure in Morocco. The water production capacity of the demonstration plant would be about 8000 m 3 /d. The objectives of pre-project study are to establish a reliable basis for a decision on a nuclear desalination plant in Morocco, using a small Chinese heating reactor and to train the Morocco experts in reactor technology and licensing aspects

The use of nuclear heat in the steel industry

International Nuclear Information System (INIS)

Coche, L.

1976-01-01

It is possible, but not easy, to use nuclear energy for steelmaking: low temperature level, and difficulty to get a continuous energy supply, are the main limiting factors. Practically, the nuclear reactor and the steel making units will not be coupled. Among the various possible systems, the most practical one for the near future consists in using nuclear heat to produce hydrogen (using natural gas or oil products as a feedstock) and electric power. Hydrogen is used to reduce iron ore in units such as Midrex, Hyl, Armco or Purofer. Steel is produced from this reduced material in electric arc furnaces. Industrial development will be slow, since economical conditions are presently pretty far from making such a process economically competitive [fr

Thermionic nuclear reactor with internal heat distribution and multiple duct cooling

Science.gov (United States)

Fisher, C.R.; Perry, L.W. Jr.

1975-11-01

A Thermionic Nuclear Reactor is described having multiple ribbon-like coolant ducts passing through the core, intertwined among the thermionic fuel elements to provide independent cooling paths. Heat pipes are disposed in the core between and adjacent to the thermionic fuel elements and the ribbon ducting, for the purpose of more uniformly distributing the heat of fission among the thermionic fuel elements and the ducts.

Studies of heat transfer having relevance to nuclear reactor containment cooling by buoyancy-driven air flow

Energy Technology Data Exchange (ETDEWEB)

Jackson, J. D.; Li, J.; Wang, J. [The Univ., of Manchester, Manchester (United Kingdom)

2003-07-01

Two separate effects experiments concerned with buoyancy-influenced convective heat transfer in vertical passages which have relevance to the problem of nuclear reactor containment cooling by means of buoyancy-driven airflow are described. A feature of each is that local values of heat transfer coefficient are determined on surfaces maintained at uniform temperature. Experimental results are presented which highlight the need for buoyancy-induced impairment of turbulent convective heat transfer to be accounted for in the design of such passive cooling systems. A strategy is presented for predicting the heat removal by combined convective and radiative heat transfer from a full scale nuclear reactor containment shell using such experimental results.

Nuclear power plant waste heat utilization

Energy Technology Data Exchange (ETDEWEB)

Ryther, J.H.; Huke, R.E.; Archer, J.C.; Price, D.R.; Jewell, W.J.; Hayes, T.D.; Witherby, H.R.

1977-09-01

The possibility of using Vermont Yankee condenser effluent for commercial food growth enhancement was examined. It was concluded that for the Vermont Yankee Nuclear Station, commercial success, both for horticulture and aquaculture endeavors, could not be assured without additional research in both areas. This is due primarily to two problems. First, the particularly low heat quality of our condenser discharge, being nominally 72 +- 2/sup 0/F; and second, to the capital intensive support systems. The capital needed for the support systems include costs of pumps, piping and controls to move the heated water to growing facilities and the costs of large, efficient heat exchangers that may be necessary to avoid regulatory difficulties due to the 1958 Delaney Amendment to the U.S. Food, Drug and Cosmetics Act. Recommendations for further work include construction of a permanent aquaculture research laboratory and a test greenhouse complex based on a greenhouse wherein a variety of heating configurations would be installed and tested. One greenhouse would be heated with biogas from an adjacent anaerobic digester thermally boosted during winter months by Vermont Yankee condenser effluent. The aquaculture laboratory would initially be dedicated to the Atlantic salmon restoration program. It appears possible to raise fingerling salmon to smolt size within 7 months using water warmed to about 60/sup 0/F. The growth rate by this technique is increased by a factor of 2 to 3. A system concept has been developed which includes an aqua-laboratory, producing 25,000 salmon smolt annually, a 4-unit greenhouse test horticulture complex and an 18,000 square foot commercial fish-rearing facility producing 100,000 pounds of wet fish (brook trout) per year. The aqualab and horticulture test complex would form the initial phase of construction. The trout-rearing facility would be delayed pending results of laboratory studies confirming its commercial viability.

Nuclear power plant waste heat utilization

International Nuclear Information System (INIS)

Ryther, J.H.; Huke, R.E.; Archer, J.C.; Price, D.R.; Jewell, W.J.; Hayes, T.D.; Witherby, H.R.

1977-09-01

The possibility of using Vermont Yankee condenser effluent for commercial food growth enhancement was examined. It was concluded that for the Vermont Yankee Nuclear Station, commercial success, both for horticulture and aquaculture endeavors, could not be assured without additional research in both areas. This is due primarily to two problems. First, the particularly low heat quality of our condenser discharge, being nominally 72 +- 2 0 F; and second, to the capital intensive support systems. The capital needed for the support systems include costs of pumps, piping and controls to move the heated water to growing facilities and the costs of large, efficient heat exchangers that may be necessary to avoid regulatory difficulties due to the 1958 Delaney Amendment to the U.S. Food, Drug and Cosmetics Act. Recommendations for further work include construction of a permanent aquaculture research laboratory and a test greenhouse complex based on a greenhouse wherein a variety of heating configurations would be installed and tested. One greenhouse would be heated with biogas from an adjacent anaerobic digester thermally boosted during winter months by Vermont Yankee condenser effluent. The aquaculture laboratory would initially be dedicated to the Atlantic salmon restoration program. It appears possible to raise fingerling salmon to smolt size within 7 months using water warmed to about 60 0 F. The growth rate by this technique is increased by a factor of 2 to 3. A system concept has been developed which includes an aqua-laboratory, producing 25,000 salmon smolt annually, a 4-unit greenhouse test horticulture complex and an 18,000 square foot commercial fish-rearing facility producing 100,000 pounds of wet fish (brook trout) per year. The aqualab and horticulture test complex would form the initial phase of construction. The trout-rearing facility would be delayed pending results of laboratory studies confirming its commercial viability

Design of SES-10 nuclear reactor for district heating

International Nuclear Information System (INIS)

Cuttler, J.M.

1991-01-01

The SES-10 units are unpressurized, pool-type nuclear reactors of 10 MW rating, designed for supplying energy to hot water district heating systems, economically and without pollution. Water for heat distribution is brought to a maximum temperature of 85 o C. Conventional heating units supplement the output from SES-10 units for peak load and during maintenance. The SES-10 is housed in a low-cost building, with a double-walled pool in the ground. A naturally circulating primary system and a pumped secondary system transport heat from the reactor to the distribution system. The unit is fully automated and easy to maintain. Because of the many active and passive safety features, it is feasible to license the SES-10 for operation in a city and easy to explain it to the public for their acceptance. The core lasts approximately 43 months at a capacity factor of 70%, and the cost of heat is expected to be 2 to 2.5 cents/kWh. (author) 8 figs

Nuclear transport of heat shock proteins in stressed cells

International Nuclear Information System (INIS)

Chughtai, Zahoor Saeed

2001-01-01

Nuclear import of proteins that are too large to passively enter the nucleus requires soluble factors, energy , and a nuclear localization signal (NLS). Nuclear protein transport can be regulated, and different forms of stress affect nucleocytoplasmic trafficking. As such, import of proteins containing a classical NLS is inhibited in starving yeast cells. In contrast, the heat shock protein hsp70 Ssa4p concentrates in nuclei upon starvation. Nuclear concentration of Ssa4p in starving cells is reversible, and transfer of nutrient-depleted cells to fresh medium induces Ssa4p nuclear export. This export reaction represents an active process that is sensitive to oxidative stress. Upon starvation, the N-terminal domain of Ssa4p mediates Ssa4p nuclear accumulation, and a short hydrophobic sequence, termed Star (for starvation), is sufficient to localize the reporter proteins green fluorescent protein or β-gaIactosidase to nuclei. To determine whether nuclear accumulation of Star-β-galactosidase depends on a specific nuclear carrier, I have analyzed its distribution in mutant yeast strains that carry a deletion of a single β-importin gene. With this assay I have identified Nmd5p as a β-importin required to concentrate Star-β-galactosidase in nuclei of stationary phase cells. (author)

Nuclear transport of heat shock proteins in stressed cells

Energy Technology Data Exchange (ETDEWEB)

Chughtai, Zahoor Saeed

2001-07-01

Nuclear import of proteins that are too large to passively enter the nucleus requires soluble factors, energy , and a nuclear localization signal (NLS). Nuclear protein transport can be regulated, and different forms of stress affect nucleocytoplasmic trafficking. As such, import of proteins containing a classical NLS is inhibited in starving yeast cells. In contrast, the heat shock protein hsp70 Ssa4p concentrates in nuclei upon starvation. Nuclear concentration of Ssa4p in starving cells is reversible, and transfer of nutrient-depleted cells to fresh medium induces Ssa4p nuclear export. This export reaction represents an active process that is sensitive to oxidative stress. Upon starvation, the N-terminal domain of Ssa4p mediates Ssa4p nuclear accumulation, and a short hydrophobic sequence, termed Star (for starvation), is sufficient to localize the reporter proteins green fluorescent protein or {beta}-gaIactosidase to nuclei. To determine whether nuclear accumulation of Star-{beta}-galactosidase depends on a specific nuclear carrier, I have analyzed its distribution in mutant yeast strains that carry a deletion of a single {beta}-importin gene. With this assay I have identified Nmd5p as a {beta}-importin required to concentrate Star-{beta}-galactosidase in nuclei of stationary phase cells. (author)

Valve arrangement for a nuclear plant residual heat removal system

International Nuclear Information System (INIS)

Fidler, G.L.; Hill, R.A.; Carrera, J.P.

1978-01-01

Disclosed is an improved valve arrangement for a two-train Residual Heat Removal System (RHRS) of a nuclear reactor plant which ensures operational integrity of the system under single failure circumstances including loss of one of two electrical power sources

Some fabrication problems in nuclear power plants heat exchanges, its detectability and implications

International Nuclear Information System (INIS)

Condessa, N.C.; Oliveira, R.

1988-01-01

On the design and manufacturing follow-up of heat-exchangers of nuclear power plants some care are took into account in order to assure a high degree of confiability allowing the heat-exchanger in operation under severe and aggressive conditions be operating during the useful life of the nuclear power plant. However, despite the care, some problems can ocurr as the ones described on this job; that, if not detected in due time could bring umpleasant problems to the component or to the system in which it is working during operation. (author) [pt

Metallic materials for heat exchanger components and highly stressed internal of HTR reactors for nuclear process heat generation

International Nuclear Information System (INIS)

1982-01-01

The programme was aimed at the development and improvement of materials for the high-temperature heat exchanger components of a process steam HTR. The materials must have high resistance to corrosion, i.e. carburisation and internal oxidation, and high long-term toughness over a wide range of temperatures. They must also meet the requirements set in the nuclear licensing procedure, i.e. resistance to cyclic stress and irradiation, non-destructive testing, etc. Initially, it was only intended to improve and qualify commercial alloys. Later on an alloy development programme was initiated in which new, non-commercial alloys were produced and modified for use in a nuclear process heat facility. Separate abstracts were prepared for 19 pays of this volume. (orig./IHOE) [de

Simplified analysis of passive residual heat removal systems for small size PWR's

International Nuclear Information System (INIS)

Botelho, D.A.

1992-02-01

The function and general objectives of a passive residual heat removal system for small size PWR's are defined. The characteristic configuration, the components and the operation modes of this system are concisely described. A preliminary conceptual specification of this system, for a small size PWR of 400 MW thermal, is made analogous to the decay heat removal system of the AP-600 reactor. It is shown by analytic models that such passive systems can dissipate 2% of nominal power within the thermal limits allowed to the reactor fuel elements. (author)

Current status of research and development for nuclear heating reactor in China

International Nuclear Information System (INIS)

Wang Dazhong; Ma Changwen; Dong Duo

1987-01-01

At present the coal is the main source for district heating in China. It results in serious problems for transportation and pollution. Nuclear district heating reactor can substitute the coal and supply the clear and ecenomic heat energy for the cities. A feasibility studies for a district heating reactor with the power of 450 MW(t) in Harbin were carried out. With cooperation of heating boilers heat demand of 1.2 million pupulation can be satisfied. 600 x 10 3 tons coal per year can be saved. The temperature of the heat grid is 130/70 deg C. The main parameters of the 450 MW(t) and 5 MW(t) heating reactors are given. The technical design, safety aspects, economic analysis and the stability of test loop are also discussed. (Liu)

Analysis of economics and market for urban nuclear heating in China

International Nuclear Information System (INIS)

Tian Jiafu; Xiao Hongchai; Jin Ziying

1987-12-01

In consideration of the heat market in northern cities, the economics of the pool-type low temperature heating reactor and the investment of urban heating system, the scale of protype reactor is chosen 200 MW, which will be able to meet the requirement of cities with more than 500 thousands population, and about 2/3 urban heat market can be met. In some areas, choosing 100 MW reactor probability is better. It is expected that the newly built apartments in North-East, North-West and North China will be about 50 million sq. meters each year and they will require some 3 GW of additional heat sources per year, and the building area incorporated into urban heating system will be 150 million sq. meters by 1990, and 500 million sq. meters by 2000. The market demand of the nuclear heat sources will increase rapidly

Prospects of heat supply from Temelin nuclear power plant

International Nuclear Information System (INIS)

Kuba, V.

1987-01-01

The possibilities are discussed of using the Temelin nuclear power plant for heat supply to a number of localities in the South Bohemian Region, to a distance of up to 34 km. Direct supply of steam and of 150/70 degC or 180/65 degC hot water is envisaged. An alternative solution has also been proposed allowing to supply steam and hot water simultaneously from 180 degC hot water with quantitative regulation. The hot water is made to expand at a pressure of 0.3 to 0.5 MPa and the low-pressure steam is compressed to a pressure of 0.9 to 1.3 MPa. This steam will be supplied to the existing heating system. The possibility was also studied of supplying Prague with heat and 180/65 degC hot water of a thermal output of up to 1,700 MW using a two-pipe heat supply line of 105 to 125 km in length. (B.S.). 2 figs

Control and Management of Small Quantity Nuclear Material (SQNM) on Safeguards

International Nuclear Information System (INIS)

Park, Jae Hwan; Shin, Byung Woo; Park, Jae Beom

2009-01-01

Small Quantity Nuclear Material (SQNM) is defined as the nuclear material that is below the amount approved in atomic energy act. SQNM generally lists depleted uranium (DU) used as a catalyst or shielding material in exposure devices in industries. The SQNM users have a duty to report information on possessing and using SQNM regularly to the government. All nuclear materials must be included in IAEA reporting lists according to safeguards agreement and additional protocol regardless of amount. However, it is difficult to investigate the status of nuclear material possessed in industries because SQNM is excepted regulation item list in atomic energy act. Most SQNM user industries are small companies so they have some problems like the loss of nuclear material after bankruptcy. Even though the damage of radiation leakage is very low, loss or careless management of nuclear material causes confusion. Thus, developing a control and management system for SQNM is essential. This paper discusses the present condition and prospect of control and management SQNM in Korea

Safety and licensing for small and medium power reactors

International Nuclear Information System (INIS)

Trauger, D.B.

1987-01-01

Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicates but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat

Safety and licensing for small and medium power reactors

International Nuclear Information System (INIS)

Trauger, D.B.

1988-01-01

Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicate but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: Nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat. (orig.)

Preliminary Study of Potential Market for Small Reactors

International Nuclear Information System (INIS)

Minato, A.; Brown, N. W.

2008-01-01

Small reactors are an energy supply for a specific purpose and oriented for a different market than large reactors. Small reactors will provide a local solution for developed and developing countries, such as, in remote areas, on small grids, or for non-electricity applications such as, district heating, seawater desalination and process heat. Single or medium sized power stations with small reactors should be compared with single fissile or renewable energy source and not be compared with large reactors. CRIEPI and LLNL have studied the business opportunities for small reactors. The small reactor concept is planned for initial use in small remote communities and in developing countries with small power distribution grid. Rapid installation and simple operation of the small plants is intended to support use in these communities without requiring development of a substantial nuclear technology infrastructure. In this study, two approaches were used in the assessment of the potential market. The first approach took a global look at the need for small nuclear plants. Then selected countries and sites were identified based on countries expressing interest in small reactors to the IAEA and consideration of sites in the US and Japan. (1) Tunisia, Mexico, Indonesia, Uruguay, Egypt and Argentina have demonstrated interest in nuclear power. Selecting one of these is dependent on political and socioeconomic factors, some of which have been identified, that require direct interaction with the countries to establish if they represent real opportunities. (2) The states of Hawaii and Alaska in the United States have high power cost and remote or island communities that may benefit from small nuclear plants. Alaska has shown greater interest in power alternatives including small than Hawaii and there is clearly less public resistance to nuclear power in Alaska. (3) The countries in Central America are actively expanding their power grids but have not demonstrated great interest

Nuclear power and heating plants in the electric power system. Part I

International Nuclear Information System (INIS)

Kalincik, L.

1975-01-01

Procedures used and results obtained in the following works are described: Incorporation of the nuclear power plants in the power system in the long term perspective; physical limitations on the WWER 440 reactor power changes during fuel campaigns; evaluation of the consumption and start-up characteristics of WWER type nuclear power plants (2x440 MWe); evaluation of refuelling campaigns distribution of nuclear power plant units with regard to comprehensive control properties of nuclear power plants; the possibilities are investigated of the utilization of the WWER type reactor for heat supply in Czechoslovakia. (author)

Nuclear process heat

International Nuclear Information System (INIS)

Barnert, H.; Hohn, H.; Schad, M.; Schwarz, D.; Singh, J.

1993-01-01

In a system for the application of high temperature heat from the HTR one must distinguish between the current generation and the use of process heat. In this respect it is important that the current can be generated by dual purpose power plants. The process heat is used as sensible heat, vaporisation heat and as chemical energy at the chemical conversion for the conversion of raw materials, the refinement of fossil primary energy carriers and finally circuit processes for the fission of water. These processes supply the market for heat, fuels, motor fuels and basic materials. Fifteen examples of HTR heat processes from various projects and programmes are presented in form of energy balances, however in a rather short way. (orig./DG) [de

Technologies for small scale wood-fueled combined heat and power systems

Energy Technology Data Exchange (ETDEWEB)

Houmann Jakobsen, H.; Houmoeller, S.; Thaaning Pedersen, L.

1998-01-01

The aim of this study is to describe and compare different technologies for small cogeneration systems (up to 2-3 MW{sub e}), based on wood as fuel. For decentralized cogeneration, i.e. for recovering energy from saw mill wood wastes or heat supply for small villages, it is vital to know the advantages and disadvantages of the different technologies. Also, for the decision-makers it is of importance to know the price levels of the different technologies. A typical obstacle for small wood cogeneration systems is the installation costs. The specific price (per kW) is usually higher than for larger plants or plants using fossil fuels. For a saw mill choosing between cogeneration and simple heat production, however, the larger installation costs are counter weighed by the sale of electricity, while the fuel consumption is the same. Whether it is profitable or not to invest in cogeneration is often hard to decide. For many years small wood cogeneration systems have been too expensive, leading to the construction of only heat producing systems due to too high price levels of small steam turbines. In recent years a great deal of effort has been put into research and developing of new technologies to replace this traditional steam turbine. Among these are: Steam engines; Stirling engines; Indirectly fired gas turbines; Pressurized down draft combustion. Along with the small scale traditional steam turbines, these technologies will be evaluated in this study. When some or all these technologies are fully developed and commercial, a strong means of reducing the strain on the environment and the greenhouse effect will be available, as the total efficiency is high (up to 90%) and wood is an energy source in balance with nature. (au) EFP-95. 19 refs.

Plate heat exchangers in the power plant industry

International Nuclear Information System (INIS)

Wersel, M.; Ridell, B.

1984-01-01

An increase in heat transfer and stability, small investment, high flexibility, easy maintenance and corrosion resistance are obtained by the design and construction of plate heat exchangers and by the introduction of the herringbone pattern. The plate heat exchanger can be used in nearly 90% of all secondary circuits in powerstations. Examples of its installation are the WYLFA, GENTILLY, RINGHALS and TVO Finland nuclear power-stations. (DG) [de

Consideration of sharing small term management under the EPC model on Nuclear power

International Nuclear Information System (INIS)

Yu Yong

2012-01-01

The special nature and importance of sharing small term under management of nuclear power EPC mode were discussed. Advantages and disadvantages of EPC mode on nuclear power.engineering is analyzed. Pointed out the management problems of owners, general contracting units, equipment suppliers and engineering construction units and other parties involved in the process of a common small term implementing. The recommendations concerning management of the owners and general contractors and subcontractors owners of nuclear power engineering inside during the management of the common aspects of sharing small term were made

CELLS OVEREXPRESSING HSP27 SHOW ACCELERATED RECOVERY FROM HEAT-INDUCED NUCLEAR-PROTEIN AGGREGATION

NARCIS (Netherlands)

KAMPINGA, HH; BRUNSTING, JF; STEGE, GJJ; KONINGS, AWT; LANDRY, J

1994-01-01

Protein denaturation/aggregation upon cell exposure to heat shock is a likely cause of cell death. in the nucleus, protein aggregation has often been correlated to inhibition of nuclear located processes and heat-induced cell killing. in Chinese hamster 023 cells made thermotolerant by a prior

Current status and development of heat exchangers for boiling water reactor nuclear power plant

International Nuclear Information System (INIS)

Uchiyama, Yoshio; Nishioka, Shuji; Ito, Shizuo

1975-01-01

More efficient and reliable operation is required for BWR heat exchangers because of nuclear plant safety and other reasons. Heat exchangers are classified into two categories of systems, the system for normal operation and the system for emergency operation. The present state and future improvement of heat exchangers are described in view of heat transfer performance, material selection, structural design, vibration, and so on. When noncondensing gas exists in vapour, heat transfer performance deteriorates, so that the heat transfer characteristics should be corrected by the adaption of venting the non condensing gas from the system. Heat transfer tubes should have high corrosion resistance to working fluid as well as high thermal conductivity, strength and economy. From that point of view, 30% cupro-nickel tubes will be replaced with 10% cupro-nickel tubes or titanium tubes though some technical development is necessary. These heat exchangers are now designed according to the MITI criteria for the technology concerning nuclear and thermal power generation, ASME Boiler and Pressure Vessel Code Sec. III and some other criteria. Most of heat transfer tube failures are caused from the tube vibration induced by working fluid flow, so that the vibration test and analysis were performed on U-tube elements. Some correction was obtained for design and fabrication techniques. (Iwase, T.)

Radwaste Management in Small Nuclear Country - National Policy and Strategy

International Nuclear Information System (INIS)

Zagar, Tomaz

2014-01-01

The lecture will briefly present the Slovene nuclear program and its legal framework focused on the radioactive waste management policy and strategy aspect. Slovenia is an example of small EU member state with small shared nuclear power program demonstrating safe, secure and efficient management of radioactive waste. Different principles of radioactive waste management will be discoursed; among others including: minimization of waste generation, the polluter pays principle, safe storage followed by final disposal and also new findings on research and development of storage, disposal and recycling of radioactive waste. (author)

Heat balance calculation and feasibility analysis for initial startup of Fuqing nuclear turbine unit with non-nuclear steam

International Nuclear Information System (INIS)

He Liu; Xiao Bo; Song Yumeng

2014-01-01

Non-nuclear steam run up compared with nuclear steam run up, can verify the design, manufacture, installation quality of the unit, at the same time shorten the follow-up duration of the entire group ready to start debugging time. In this paper, starting from the first law of thermodynamics, Analyzed Heat balance Calculation and Feasibility analysis for Initial startup of Fuqing nuclear Turbine unit with Non-nuclear steam, By the above calculation, to the system requirements and device status on the basis of technical specifications, confirmed the feasibility of Non-nuclear steam running up in theory. After the implementation of the Non-nuclear turn of Fuqing unit, confirmed the results fit with the actual process. In summary, the Initial startup of Fuqing turbine unit with Non-nuclear steam is feasible. (authors)

The nuclear battery

International Nuclear Information System (INIS)

Kozier, K.S.; Rosinger, H.E.

1988-01-01

This paper reviews the evolution and present status of an Atomic Energy of Canada Limited program to develop a small, solid-state, passively cooled reactor power supply known as the Nuclear Battery. Key technical features of the Nuclear Battery reactor core include a heat-pipe primary heat transport system, graphite neutron moderator, low-enriched uranium TRISO coated-particle fuel and the use of burnable poisons for long-term reactivity control. An external secondary heat transport system extracts useful heat energy, which may be converted into electricity in an organic Rankine cycle engine or used to produce high-pressure steam. The present reference design is capable of producing about 2400 kW(t) (about 600 kW(e) net) for 15 full-power years. Technical and safety features are described along with recent progress in component hardware development programs and market assessment work. 19 refs

Nuclear heating measurements by in-pile calorimetry: prospective works for a microsensor design

Energy Technology Data Exchange (ETDEWEB)

Reynard-Carette, C.; Carette, M.; Aguir, K.; Bendahan, M.; Fiorido, T. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Lyoussi, A.; Fourmentel, D.; Villard, J.F. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 (France); Barthes, M.; Lanzetta, F.; Layes, G.; Vives, S. [FEMTO-ST, UMR 6174, Departement ENERGIE, Universite de Franche-Comte, 90000, Belfort (France)

2015-07-01

Since 2009 works have been performed in the framework of joint research programs between CEA and Aix-Marseille University. The main aim of these programs is to design and develop in-pile instrumentations, advanced calibration procedure and accurate measurement methods in particular for the new Material Testing Reactor (MTR) under construction in the South of France: Jules Horowitz Reactor (JHR). One major sensor is a specific radiometric calorimeter, which was studied out-of-pile from a thermal point of view and in-pile during irradiation campaigns. This sensor type is dedicated to measurements of nuclear heating (energy deposition rate per mass unit induced by interactions between nuclear rays and matter) inside experimental channels of MTRs. This kind of in-pile calorimeter corresponds to heat flux calorimeter exchanging with the external cooling fluid. This thermal running mode allows the establishment of steady thermal conditions inside the sensor to carry out online continuous measurements inside the reactor (core or reflector). Two main types of calorimeters exist. The first type consists of a single cell calorimeter. It is divided into a sample of material to be tested and a jacket instrumented with two thermocouples or a single thermocouple (Gamma Thermometer). The second, called a differential calorimeter, is composed of two superposed twin cells (a measurement cell containing a sample of material, and a reference cell to remove the heating of the cell body) instrumented with four thermocouples and two electrical heaters. Contrary to a single-cell calorimeter, a differential calorimeter allows the compensation of the parasite nuclear heating of the sensor body or jacket. Moreover, it possesses interesting advantages: thanks to the heaters embedded in the cells, three different measurement methods can be applied during irradiations to quantify nuclear heating. The first one is based on the use of out-of-pile calibration curves obtained by generating a heat

Time evolution simulation of heat removal in a small water tank by natural convection

Energy Technology Data Exchange (ETDEWEB)

Freitas, Carlos Alberto de, E-mail: carlos.freitas1950@hotmail.com [Instituto Federal do Rio de Janeiro (IFRJ), Nilopolis, RJ (Brazil); Jachic, Joao; Moreira, Maria de Lourdes, E-mail: jjachic@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2013-07-01

One of the cooling modes for any source of heat such as in a shutdown nuclear core is the natural convection. The design specifications of any cooling pool can only be done when the removal heat rate and the corresponding mass flow rate is reasonably established. In our simulation scheme, we assumed that the body forces acting in the cubic water cell are: the weight, the drag force and the integrated pressure forces on the horizontal surfaces, the viscosity shear forces on the vertical surfaces and also a special viscosity drag force due to the mass dislocation along a Bernoulli type current tube outside the motive region. For a suitable time step, the uprising convection velocity is determined by an implicit and also by an explicit solution algorithm. The resulting differential equation depends on updating specific mass, dynamic viscosity and constant pressure heat coefficient with the last known temperature in the cell that absorbed heat. Numerical calculation software was performed using MATLAB’s technical computing language and then applied for a heat generation plate simulating a spent fuel assembler from a shutdown nuclear core. The results show time evolution of convection, terminal velocity and water temperature distribution. Pool dimension as well as pool level decrement are also determined for various air exhausting system conditions and heat rate of the spent fuel plate being cooled. (author)

Time evolution simulation of heat removal in a small water tank by natural convection

International Nuclear Information System (INIS)

Freitas, Carlos Alberto de; Jachic, Joao; Moreira, Maria de Lourdes

2013-01-01

One of the cooling modes for any source of heat such as in a shutdown nuclear core is the natural convection. The design specifications of any cooling pool can only be done when the removal heat rate and the corresponding mass flow rate is reasonably established. In our simulation scheme, we assumed that the body forces acting in the cubic water cell are: the weight, the drag force and the integrated pressure forces on the horizontal surfaces, the viscosity shear forces on the vertical surfaces and also a special viscosity drag force due to the mass dislocation along a Bernoulli type current tube outside the motive region. For a suitable time step, the uprising convection velocity is determined by an implicit and also by an explicit solution algorithm. The resulting differential equation depends on updating specific mass, dynamic viscosity and constant pressure heat coefficient with the last known temperature in the cell that absorbed heat. Numerical calculation software was performed using MATLAB’s technical computing language and then applied for a heat generation plate simulating a spent fuel assembler from a shutdown nuclear core. The results show time evolution of convection, terminal velocity and water temperature distribution. Pool dimension as well as pool level decrement are also determined for various air exhausting system conditions and heat rate of the spent fuel plate being cooled. (author)

Participation of nuclear power plants in variable operation regimes under conditions of combined electric power and heat generation

International Nuclear Information System (INIS)

Rydzi, S.

1988-01-01

The incorporation of nuclear power units in the control of the output of an electric power system is affected by technical and economic factors as well as by the manner of heat take-off from the nuclear power unit for heating purposes. The effect was therefore studied of the technological solution of converting the heat output of WWER-440 units to operating parameters of turbines in nonrated regimes of operation. Some results of the study are graphically represented. An analysis was also made of limitations preventing WWER-440 units from supplying heat with regard to their incorporation in the electric power transmission system. The results show that using nuclear power units for district heating will in the future strictly determine the seasonal shut-down of nuclear units for fuel exchange and overhauls. This could interfere with the considered concept of the 1.5 year duty time of WWER-440 reactors. With regard to the economy of operation of the nuclear power system and reduced demands on weekend unloading it will be necessary to incorporate in the power system pumped-storage power plants with one-week pumped-storage systems. (Z.M.). 5 figs., 2 tabs., 6 refs

District heating with SLOWPOKE energy systems

International Nuclear Information System (INIS)

Lynch, G.F.

1988-03-01

The SLOWPOKE Energy System, a benign nuclear heat source designed to supply 10 thermal megawatts in the form of hot water for local heating systems in buildings and institutions, is at the forefront of these developments. A demonstration unit has been constructed in Canada and is currently undergoing an extensive test program. Because the nuclear heat source is small, operates at atmospheric pressure, and produces hot water below 100 degrees Celcius, intrinsic safety features will permit minimum operator attention and allow the heat source to be located close to the load and hence to people. In this way, a SLOWPOKE Energy System can be considered much like the oil- or coal-fired furnace it is designed to replace. The low capital investment requirements, coupled with a high degree of localization, even for the first unit, are seen as attractive features for the implementation of SLOWPOKE Energy Systems in many countries

Market Potential Estimation of a Small and Medium Size Reactor

International Nuclear Information System (INIS)

Oh, K. B.; Yang, M. H.; Lee, M. K.; Chung, W. S.; Kim, H. J.; Kim, S. S.; Lee, B. W.; Ryu, J. S.; Juhn, P. E.

2004-12-01

Technically, nuclear reactors which produce energy in the form of heat can supply energy products other than electricity, including district heat, process heat, potable water etc. Currently non-civil uses of nuclear energy are very limited civil applications except for a military nuclear powered ship and an energy usage in isolated areas. However, the future global environment and an energy resources scarcity could promote a significant usage of non-electrical applications on an industrial scale. Considering these situations, this report analyzed the following: (1) Worldwide non-electrical application of a small and medium sized nuclear reactor - survey the situation of the current technical applications, - survey the global market potential estimation for various applications (2) Technical cooperation potential in several countries - identify necessary conditions for nuclear cooperation, - select candidate countries: Morocco, UAE, Indonesia, Chile and Vietnam, - survey the energy and water situation, - survey the legal and international regime infrastructure

Primary and secondary structure of U8 small nuclear RNA

International Nuclear Information System (INIS)

Reddy, R.; Henning, D.; Busch, H.

1985-01-01

U8 small nuclear RNA is a new, capped, 140 nucleotides long RNA species found in Novikoff hepatoma cells. Its sequence is: m3GpppAmUmCGUCAGGA GGUUAAUCCU UACCUGUCCC UCCUUUCGGA GGGCAGAUAG AAAAUGAUGA UUGGAGCUUG CAUGAUCUGC UGAUUAUAGC AUUUCCGUGU AAUCAGGACC UGACAACAUC CUGAUUGCUU CUAUCUGAUUOH. This RNA is present in approximately 25,000 copies/cell, and it is enriched in nucleolar preparations. Like U1, U2, U4/U6, and U5 RNAs, U8 RNA was also present as a ribonucleoprotein associated with the Sm antigen. The rat U8 RNA was highly homologous (greater than 90%) to a recently characterized 5.4 S RNA from mouse cells infected with spleen focus-forming virus. In addition to the U8 RNA, three other U small nuclear RNAs were found in anti-Sm antibody immunoprecipitates from labeled rat and HeLa cells. Each of these contained a m3GpppAm cap structure; their apparent chain lengths were 60, 130, and 65 nucleotides. These U small nuclear RNAs are designated U7, U9, and U10 RNAs, respectively

On the optimal design of shell and tube heat exchanger for nuclear applications

International Nuclear Information System (INIS)

Abd Rabbo, F.M.M.; Fatb Allab, A.; El-Fawal, M.

1997-01-01

In nuclear industry, heat exchanger plays an important role in the transfer of heat from reactor core, where heat is generated, to the ultimate heat sink UHS, and then is dissipated. The actual design of heat exchanger not only relies on thermohydraulic considerations but also on economical aspects and radiological safety considerations. For optimal design of heat exchanger for a specific application a compromise should be made for determining the important factors affecting the design. In this paper, an optimization model is presented for shell and tube heat exchanger, which could be considered as a tool for computer aided design. A case study is presented to explore the present adopted model. 3 figs

Large-Scale Combined Heat and Power (CHP) Generation at Loviisa Nuclear Power Plant Unit 3

International Nuclear Information System (INIS)

Bergroth, N.

2010-01-01

Fortum has applied for a Decision in Principle concerning the construction of a new nuclear power plant unit (Loviisa 3) ranging from 2800-4600 MWth at its site located at the southern coast of Finland. An attractive alternative investigated is a co-generation plant designed for large-scale district heat generation for the Helsinki metropolitan area that is located approximately 75 km west of the site. The starting point is that the district heat generation capacity of 3 unit would be around 1 000 MWth.The possibility of generating district heat for the metropolitan area by Loviisa's two existing nuclear power plant units was investigated back in the 1980s, but it proved unpractical at the time. With the growing concern of the climate change and the subsequent requirements on heat and power generation, the idea is much more attractive today, when recognising its potential to decrease Finland's carbon dioxide emissions significantly. Currently the district heat generation in metropolitan area is based on coal and natural gas, producing some five to seven million tonnes of carbon dioxide emissions annually. Large-scale combined heat and power (CHP) generation at the 3 unit could cut this figure by up to four million tonnes. This would decrease carbon dioxide emissions by as much as six percent. In addition, large-scale CHP generation would increase the overall efficiency of the new unit significantly and hence, reduce the environmental impact on the local marine environment by cutting heat discharges into the Gulf of Nuclear energy has been used for district heating in several countries both in dedicated nuclear heating plants and in CHP generation plants. However, the heat generation capacity is usually rather limited, maximum being around 250 MWth per unit. Set against this, the 3 CHP concept is much more ambitious, not only because of the much larger heat generation output envisaged, but also because the district heating water would have to be transported over a

Heat supply of the town of Trnava from EBO nuclear plant

International Nuclear Information System (INIS)

Kovarik, Z.

4 variants of central heat supply from the V-2 Bohunice nuclear power plant were considered. Three included the supply of hot water while the fourth envisaged the supply of steam. The variants of hot water supply differed in the proportion of heat supply during load peak, the range being from 75% to 88% to 100%. The system suggested considers hot water with a maximum overpressure of 2.5 MPa and temperature of 150/70 degC, with possible temporary decrease to 130/70 degC. The estimated power of heat exchange stations is 240 MW at a temperature gradient of 150/70 degC and 180 MW at a temperature gradient 130/70 degC. The location, design and control of the heat supply system and the specifications of heat generation facilities are given. (J.P.)

Small reactor operating mode

International Nuclear Information System (INIS)

Snell, V.G.

1997-01-01

There is a potential need for small reactors in the future for applications such as district heating, electricity production at remote sites, and desalination. Nuclear power can provide these at low cost and with insignificant pollution. The economies required by the small scale application, and/or the remote location, require a review of the size and location of the operating staff. Current concepts range all the way from reactors which are fully automatic, and need no local attention for days or weeks, to those with reduced local staff. In general the less dependent a reactor is on local human intervention, the greater its dependence on intrinsic safety features such as passive decay heat removal, low-stored energy and limited reactivity speed and depth in the control systems. A case study of the design and licensing of the SLOWPOKE Energy System heating reactor is presented. (author)

Heat transfer with water in forced convection without boiling in small diameter tubes

International Nuclear Information System (INIS)

Ricque, Roger; Siboul, Roger

1969-01-01

This note presents the measurements performed for the establishment of an empirical heat transfer law for water in forced convection without boiling in small diameter tubes (2 and 4 mm), with high flow velocity and strong heat flux, and for relatively low fluid temperatures. A correlation of experimental points is obtained with a very small maximum dispersion: Nu fl = 0,0092 Re fl 0,88 Pr 0,5 (μ fl /μ p ) 0,14 . A correlation for the fiction coefficient is also presented [fr

Old foes, new understandings: nuclear entry of small non-enveloped DNA viruses.

Science.gov (United States)

Fay, Nikta; Panté, Nelly

2015-06-01

The nuclear import of viral genomes is an important step of the infectious cycle for viruses that replicate in the nucleus of their host cells. Although most viruses use the cellular nuclear import machinery or some components of this machinery, others have developed sophisticated ways to reach the nucleus. Some of these have been known for some time; however, recent studies have changed our understanding of how some non-enveloped DNA viruses access the nucleus. For example, parvoviruses enter the nucleus through small disruptions of the nuclear membranes and nuclear lamina, and adenovirus tugs at the nuclear pore complex, using kinesin-1, to disassemble their capsids and deliver viral proteins and genomes into the nucleus. Here we review recent findings of the nuclear import strategies of three small non-enveloped DNA viruses, including adenovirus, parvovirus, and the polyomavirus simian virus 40. Copyright © 2015 Elsevier B.V. All rights reserved.

Heat Loads Due To Small Penetrations In Multilayer Insulation Blankets

Science.gov (United States)

Johnson, W. L.; Heckle, K. W.; E Fesmire, J.

2017-12-01

The main penetrations (supports and piping) through multilayer insulation systems for cryogenic tanks have been previously addressed by heat flow measurements. Smaller penetrations due to fasteners and attachments are now experimentally investigated. The use of small pins or plastic garment tag fasteners to ease the handling and construction of multilayer insulation (MLI) blankets goes back many years. While it has long been understood that penetrations and other discontinuities degrade the performance of the MLI blanket, quantification of this degradation has generally been lumped into gross performance multipliers (often called degradation factors or scale factors). Small penetrations contribute both solid conduction and radiation heat transfer paths through the blanket. The conduction is down the stem of the structural element itself while the radiation is through the hole formed during installation of the pin or fastener. Analytical models were developed in conjunction with MLI perforation theory and Fourier’s Law. Results of the analytical models are compared to experimental testing performed on a 10 layer MLI blanket with approximately 50 small plastic pins penetrating the test specimen. The pins were installed at ∼76-mm spacing inches in both directions to minimize the compounding of thermal effects due to localized compression or lateral heat transfer. The testing was performed using a liquid nitrogen boil-off calorimeter (Cryostat-100) with the standard boundary temperatures of 293 K and 78 K. Results show that the added radiation through the holes is much more significant than the conduction down the fastener. The results are shown to be in agreement with radiation theory for perforated films.

Potential advantages and disadvantages of sequentially building small nuclear units instead of a large nuclear plant

International Nuclear Information System (INIS)

Feretic, D.; Cavlina, N.; Grgic, D.

2008-01-01

Renewal of nuclear power programs in countries with modest electricity consumptions and weak electrical grid interconnections has raised the question of optimal nuclear power plants sizes for such countries. The same question would be also valid for isolated or weakly connected regions within a large country. Building large size nuclear power plant could be prevented by technical or financial limits. Research programs have been initiated in the International Atomic Energy Agency and in the USA (within the framework of the Global Nuclear Energy Partnership (GNEP) program) with the aim to inspect under which circumstances small and medium reactors could be the preferred option compared to large nuclear plants. The economy of scale is a clear advantage of large plants. This paper compares, by using probabilistic methods, the net cash flow of large and medium size plants, taking as example a large nuclear plant (around 1200 MW) and four sequentially built smaller plants (300 MW). Potential advantages and disadvantageous of both options have been considered. Main advantages of the sequential construction of several identical small units could be the reduced investor risk and reduced investment costs due to the learning effect. This analysis is a part of studies for the Croatian power generating system development. (orig.)

Preliminary study on high temperature heat exchanger for nuclear steel making

Energy Technology Data Exchange (ETDEWEB)

Mori, Y [Tokyo Inst. of Tech. (Japan); Ikegami, H

1975-03-01

In the high temperature heat exchanger as well as the steam reformer, several technical problems should be solved before realizing a nuclear plant complex for iron and steel making. Research has been carried out on heat exchanger between helium and steam, hydrogen permeation through super alloys, hydrogen removal using a titanium sponge, and creep and carburization performance of super alloys. The primary coolant used is helium having a pressure of approximately 12 kg/cm/sup 2/G and a temperature of approximately 1100/sup 0/C measured at the inlet of the high temperature heat exchanger, i.e., the test section. Steam, hydrogen and carbon monoxide are used as secondary coolants.

Heat transfer coefficient testing in nuclear fuel rod bundles with mixing vane grids

International Nuclear Information System (INIS)

Conner, Michael E.; Smith, L. David III; Holloway, Mary V.; Beasley, Donald E.

2005-01-01

An air heat transfer test facility was developed to test the heat transfer downstream of support grids in simulated PWR nuclear fuel rod bundles. The goal of this testing is to study the single-phase heat transfer coefficients downstream of grids with mixing vanes in a square-pitch rod bundle. The technique developed utilizes fully-heated grid spans and a specially designed thermocouple holder that can be moved axially down the rod bundle and aximuthally within a test rod. From this testing, the axial and aximuthally varying heat transfer coefficient can be determined. Different grid designs are tested and compared to determine the heat transfer enhancement associated with key grid features such as mixing vanes. (author)

Enhancement of nuclear heat transfer in a typical pressurized water reactor by new spacer grids

International Nuclear Information System (INIS)

Nazifi, M.; Nematollahi, M.

2007-01-01

The fuel element geometry typically used in nuclear reactor is rod bundle whose rod-to-rod clearance is maintained by grid spacer. The heat generated in the rod by nuclear reaction is removed by coolant, usually in turbulent flow. The coolant moves axially through the subchannels. Fuel spacer grid affects the coolant flow distribution in a fuel rod bundle, and so spacer geometry has a strong influence on a bundle's thermal-hydraulic characteristics such as critical heat flux and pressure drop. An understanding of the detailed structure of the turbulent flow and heat transfer in the rod bundle, used especially as nuclear fuel elements, is of major interest to the nuclear power industry for their safe and reliable operation. The flow mixing devices on grid spacer would enhance the mixing rate between sub-channels and promote the turbulence in subchannel. The present study evaluates the effects of mixing vane shape on flow structure and heat transfer downstream of mixing vane in a sub-channel of fuel assembly, by obtaining velocity and pressure fields, turbulent intensity, flow mixing factors, heat transfer coefficient and friction factor using three-dimensional RANS analysis. Six new shapes mixing vane designed by the authors, are simulated numerically to evaluate the performance in enhancing the heat transfer, in comparison with commercialized split vane. Standard K-epsilon model are used as a turbulence closure model and periodic and symmetry condition are set as boundary conditions. The capability of the model to predict the coolant flow distribution inside rod bundles is shown and discussed on the base of comparison with experimental data for a variety of geometrical and Reynolds number conditions. It is conformed that the turbulence in the sub-channel was significantly promoted by spacer and mixing devices but rapidly decreased to a fully developed level approximately 10 time of hydraulic diameter downstream of the top of spacer. Ring type mixer showed a high

Small-scale automated biomass energy heating systems: a viable option for remote Canadian communities?

Energy Technology Data Exchange (ETDEWEB)

McCallum, B. [Canadian Forest Service, Ottawa, ON (Canada). Industry, Economics and Programs Branch

1997-12-31

The potential benefits of wood energy (forest biomass) for space heating in Canada`s remote communities was discussed. Diesel fuel and heating oil must be transported into these communities to produce electricity and to heat large public buildings. Below the treeline, roundwood is often used to heat private homes. The move toward environmentally sustainable development has focussed much attention on renewable energy technologies such as biomass energy, (i.e. any form of energy derived from plant or animal materials). Wood is the most readily available biomass fuel in remote communities. Woodchips and sawmill waste can be burned in automated biomass heating systems which provide a convenient way to use low-grade wood to heat large buildings or groups of buildings which would not be feasible to heat with roundwood. It was shown that one cord of spruce can produce 1.5 tonnes of woodchips to ultimately displace 300 litres of heating oil. A description of a small-commercial and small-industrial biomass system was presented. The benefits of biomass were described as: (1) direct savings compared to high-cost oil heat, (2) increased circulation of energy dollars inside the community, and (3) employment opportunities in harvesting, processing and operating biomass systems. A steady supply of good quality woodchips to the heating plant must be ensured. 1 ref., 3 figs.

The entropy problem of the decentralized solar and nuclear heat generation

International Nuclear Information System (INIS)

Seifritz, W.

1984-01-01

Parallel to the energy fluxes the entropy fluxes of decentralized hot-water systems based on solar collectors coupled with an electrical auxiliary heating installation are also deduced. As an important result the fact emerges that this kind of solar energy has to remain very restricted, not only for quantitative-energetic reasons, but also for entropy ones, and that a solar hot-water system will always have to rely on an energy system of low entropy. In contrast to this, the provision of heat for space heating sector with the help of the 'nuclear short-distance concept', which practically does not need any external energy, is not subject to these restrictions. This concept is introduced briefly, as well as the heat prices which presumably can be achieved by it. Concluding comments summarize the reasons once again that speak against the installation of a decentralized solar heat supply system. (orig.) [de

Calorimeter measures high nuclear heating rates and their gradients across a reactor test hole

Science.gov (United States)

Burwell, D.; Coombe, J. R.; Mc Bride, J.

1970-01-01

Pedestal-type calorimeter measures gamma-ray heating rates from 0.5 to 7.0 watts per gram of aluminum. Nuclear heating rate is a function of cylinder temperature change, measured by four chromel-alumel thermocouples attached to the calorimeter, and known thermoconductivity of the tested material.

Study of two control rods of a district heating nuclear plant

International Nuclear Information System (INIS)

Martinez, J.M.

1979-01-01

This paper broaches the study of the control rods to ensure a convenient working during load following of the nuclear reactor THERMOS. The mathematical model is descriptive of the whole of the nuclear plant (point model for the core and the heat balances). Two power control are studied. The first, like PWR, is a program for the mean temperature of primary water. The second takes into account the structure of the plant and is described by a schedule of powers [fr

Selection of heat disposal methods for a Hanford Nuclear Energy Center

International Nuclear Information System (INIS)

Young, J.R.; Kannberg, L.D.; Ramsdell, J.V.; Rickard, W.H.; Watson, D.G.

1976-06-01

Selection of the best method for disposal of the waste heat from a large power generation center requires a comprehensive comparison of the costs and environmental effects. The objective is to identify the heat dissipation method with the minimum total economic and environmental cost. A 20 reactor HNEC will dissipate about 50,000 MWt of waste heat; a 40 reactor HNEC would release about 100,000 MWt. This is a much larger discharge of heat than has occurred from other concentrated industrial facilities and consequently a special analysis is required to determine the permissibility of such a large heat disposal and the best methods of disposal. It is possible that some methods of disposal will not be permissible because of excessive environmental effects or that the optimum disposal method may include a combination of several methods. A preliminary analysis is presented of the Hanford Nuclear Energy Center heat disposal problem to determine the best methods for disposal and any obvious limitations on the amount of heat that can be released. The analysis is based, in part, on information from an interim conceptual study, a heat sink management analysis, and a meteorological analysis

Small-Scale Combined Heat and Power Plants Using Biofuels

Energy Technology Data Exchange (ETDEWEB)

Salomon-Popa, Marianne [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Energy Technology

2002-11-01

In this time period where energy supply and climate change are of special concern, biomass-based fuels have attracted much interest due to their plentiful supply and favorable environmental characteristics (if properly managed). The effective capture and continued sustainability of this renewable resource requires a new generation of biomass power plants with high fuel energy conversion. At the same time, deregulation of the electricity market offers new opportunities for small-scale power plants in a decentralized scheme. These two important factors have opened up possibilities for small-scale combined heat and power (CHP) plants based on biofuels. The objective of this pre-study is to assess the possibilities and technical limitations for increased efficiency and energy utilization of biofuels in small size plants (approximately 10 MWe or lower). Various energy conversion technologies are considered and proven concepts for large-scale fossil fuel plants are an especially important area. An analysis has been made to identify the problems, technical limitations and different possibilities as recognized in the literature. Beyond published results, a qualitative survey was conducted to gain first-hand, current knowledge from experts in the field. At best, the survey results together with the results of personal interviews and a workshop on the role of small-scale plants in distributed generation will serve a guideline for future project directions and ideas. Conventional and novel technologies are included in the survey such as Stirling engines, combustion engines, gas turbines, steam turbines, steam motors, fuel cells and other novel technologies/cycles for biofuels. State-of-the-art heat and power plants will be identified to clarify of the advantages and disadvantages as well as possible obstacles for their implementation.

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

A new small HTGR power plant concept with inherently safe features--An engineering and economic challenge

International Nuclear Information System (INIS)

McDonald, C.F.; Sonn, D.L.

1983-01-01

This paper outlines a small nuclear plant concept which is not meant to replace the large nuclear power plants that will continue to be needed by the industrialized nations, but rather recognizes the needs of the smaller energy user, both for special applications in the US and for the developing nations. The small High-Temperature Gas-Cooled Reactor (HTGR), whose introduction will be very dependent on market forces, represents only one approach to meet these needs. The design of a small power plant that could be inherently safer and that might have costs less than those indicated by the traditional reverse-economy-of-scale effect is discussed. Topics considered include power plant economics, the small steam cycle HTGR thermodynamic cycle, the reactor nuclear heat source layout, the reactor heat removal system (main loop cooling, a vessel cooling system with reactor pressurized, vessel cooling system with reactor depressurized), safety considerations, investment risk protection, the technology base, and applications for the small HTGR plant concept

Research of management information system of radiation protection for low temperature nuclear heating reactor

International Nuclear Information System (INIS)

Bai Hongtao; Wang Jiaying; Wu Manxue

2001-01-01

Management information system of radiation protection for low temperature reactor uses computer to manage the data of the low temperature nuclear heating reactor radiation monitoring, it saves the data from the front real-time radiation monitoring system, comparing these data with historical data to give the consequence. Also, the system provides some picture in order to show space information at need. The system, based on Microsoft Access 97, consists of nine parts, including radiation dose, environmental data, meteorological data and so on. The system will have value in safely operation of the low temperature nuclear heating reactor

Design of an Experimental Facility for Passive Heat Removal in Advanced Nuclear Reactors

Science.gov (United States)

Bersano, Andrea

With reference to innovative heat exchangers to be used in passive safety system of Gen- eration IV nuclear reactors and Small Modular Reactors it is necessary to study the natural circulation and the efficiency of heat removal systems. Especially in safety systems, as the decay heat removal system of many reactors, it is increasing the use of passive components in order to improve their availability and reliability during possible accidental scenarios, reducing the need of human intervention. Many of these systems are based on natural circulation, so they require an intense analysis due to the possible instability of the related phenomena. The aim of this thesis work is to build a scaled facility which can reproduce, in a simplified way, the decay heat removal system (DHR2) of the lead-cooled fast reactor ALFRED and, in particular, the bayonet heat exchanger, which transfers heat from lead to water. Given the thermal power to be removed, the natural circulation flow rate and the pressure drops will be studied both experimentally and numerically using the code RELAP5 3D. The first phase of preliminary analysis and project includes: the calculations to design the heat source and heat sink, the choice of materials and components and CAD drawings of the facility. After that, the numerical study is performed using the thermal-hydraulic code RELAP5 3D in order to simulate the behavior of the system. The purpose is to run pretest simulations of the facility to optimize the dimensioning setting the operative parameters (temperature, pressure, etc.) and to chose the most adequate measurement devices. The model of the system is continually developed to better simulate the system studied. High attention is dedicated to the control logic of the system to obtain acceptable results. The initial experimental tests phase consists in cold zero power tests of the facility in order to characterize and to calibrate the pressure drops. In future works the experimental results will be

Strain components of nuclear-reactor-type concretes during first heat cycle

International Nuclear Information System (INIS)

Khoury, G.A.

1995-01-01

Strains of three advanced-gas-cooled-reactor-type nuclear reactor concretes were measured during the first heat cycle and their relative thermal stability determined. It was possible to isolate for the first time the shrinkage component for the period during heating. Predictions of the residual strains for the loaded specimens can be made by simple superposition of creep and shrinkage components up to a certain critical temperature, which for basalt concrete is about 500 C and for limestone concrete is about 200-300 C. Above the critical temperature, an expansive ''cracking'' strain component is present. It is shown that the strain behaviour of concrete provides a sensitive indication of its thermal stability during heating and subsequent cooling. (orig.)

High-Efficiency Small-Scale Combined Heat and Power Organic Binary Rankine Cycles

Directory of Open Access Journals (Sweden)

Costante Mario Invernizzi

2018-04-01

Full Text Available Small-CHP (Combined Heat and Power systems are generally considered a valuable technological option to the conventional boilers, in a technology developed context. If small-CHP systems are associated with the use of renewable energies (biomass, for example they could play an important role in distributed generation even in developing countries or, in any case, where there are no extensive electricity networks. Traditionally the considered heat engines for micro- or mini-CHP are: the gas engine, the gas turbine (with internal combustion, the steam engine, engine working according to the Stirling and to the Rankine cycles, the last with organic fluids. In principle, also fuel cells could be used. In this paper, we focus on small size Rankine cycles (10–15 k W with organic working fluids. The assumed heat source is hot combustion gases at high temperature (900–950 ∘ C and we assume to use only single stages axial turbines. The need to work at high temperatures, limits the choice of the right organic working fluids. The calculation results show the limitation in the performances of simple cycles and suggest the opportunity to resort to complex (binary cycle configurations to achieve high net conversion efficiencies (15–16%.

Thermoacoustic sensor for nuclear fuel temperaturemonitoring and heat transfer enhancement

Energy Technology Data Exchange (ETDEWEB)

James A. Smith; Dale K. Kotter; Randall A. Alli; Steven L. Garrett

2013-05-01

A new acoustical sensing system for the nuclear power industry has been developed at The Pennsylvania State University in collaboration with Idaho National Laboratories. This sensor uses the high temperatures of nuclear fuel to convert a nuclear fuel rod into a standing-wave thermoacoustic engine. When a standing wave is generated, the sound wave within the fuel rod will be propagated, by acoustic radiation, through the cooling fluid within the reactor or spent fuel pool and can be monitored a remote location external to the reactor. The frequency of the sound can be correlated to an effective temperature of either the fuel or the surrounding coolant. We will present results for a thermoacoustic resonator built into a Nitonic-60 (stainless steel) fuel rod that requires only one passive component and no heat exchangers.

Atomistic Simulations of Small-scale Materials Tests of Nuclear Materials

International Nuclear Information System (INIS)

Shin, Chan Sun; Jin, Hyung Ha; Kwon, Jun Hyun

2012-01-01

Degradation of materials properties under neutron irradiation is one of the key issues affecting the lifetime of nuclear reactors. Evaluating the property changes of materials due to irradiations and understanding the role of microstructural changes on mechanical properties are required for ensuring reliable and safe operation of a nuclear reactor. However, high dose of neuron irradiation capabilities are rather limited and it is difficult to discriminate various factors affecting the property changes of materials. Ion beam irradiation can be used to investigate radiation damage to materials in a controlled way, but has the main limitation of small penetration depth in the length scale of micro meters. Over the past decade, the interest in the investigations of size-dependent mechanical properties has promoted the development of various small-scale materials tests, e.g. nanoindentation and micro/nano-pillar compression tests. Small-scale materials tests can address the issue of the limitation of small penetration depth of ion irradiation. In this paper, we present small-scale materials tests (experiments and simulation) which are applied to study the size and irradiation effects on mechanical properties. We have performed molecular dynamics simulations of nanoindentation and nanopillar compression tests. These atomistic simulations are expected to significantly contribute to the investigation of the fundamental deformation mechanism of small scale irradiated materials

Fuel production from coal by the Mobil Oil process using nuclear high-temperature process heat

International Nuclear Information System (INIS)

Hoffmann, G.

1982-01-01

Two processes for the production of liquid hydrocarbons are presented: Direct conversion of coal into fuel (coal hydrogenation) and indirect conversion of coal into fuel (syngas production, methanol synthesis, Mobil Oil process). Both processes have several variants in which nuclear process heat may be used; in most cases, the nuclear heat is introduced in the gas production stage. The following gas production processes are compared: LURGI coal gasification process; steam reformer methanation, with and without coal hydrogasification and steam gasification of coal. (orig./EF) [de

Import and export of small quantities of nuclear materials

International Nuclear Information System (INIS)

Grenier, M.

1986-06-01

Administrative procedures for import export of nuclear materials are specific for each country. In France regulations are reviewed for small quantities, lower threshold, in some cases, allows a simplified procedure, however thresholds are not the same in the different texts (and for one of them, concerning proliferation, is zero). It is obvious that regulations are necessary even for small quantities but national and international threshold should be harmonized [fr

Nuclear DVCS at small x using color-dipole phenomenology

International Nuclear Information System (INIS)

Machado, Magno V.T.

2009-01-01

Using the high-energy color-dipole formalism, we study the coherent and incoherent nuclear DVCS process, γ * A→γX, in the small-x regime. We consider simple models for the elementary dipole-hadron scattering amplitude that capture the main features of the dependence on atomic number A, on energy and on momentum transfer t. Using the amplitudes obtained we make predictions for the nuclear DVCS cross section at the photon level in collider kinematics. (orig.)

Comparison of the Thermal Response of Two Calorimetric Cells Dedicated to Nuclear Heating Measurements during Calibration

International Nuclear Information System (INIS)

Brun, J.; Reynard, C.; De-Vita, C.; Carette, M.; Muraglia, M.; Lyoussi, A.; Fourmentel, D.; Guimbal, P.; Villard, J-F.

2013-06-01

Nuclear heating is a key parameter which contributes to the thermal design and the quality of in-pile experiments performed in Material Testing Reactors (MTRs) for the study of nuclear materials and fuels under irradiation. Nuclear heating is typically measured in MTRs by radiometric calorimeters. However this kind of sensor has to be suited and improved in perspective of the new experimental conditions inside the channels of Jules Horowitz Reactor (JHR). In this paper, we study the responses of two non adiabatic differential calorimeter cells having the same geometric design, but different dimensions. These experimental works are carried out during a preliminary out-of-pile calibration operating procedure of these sensors which consists in simulating the sample heating by Joule effect. The influence of the imposed electrical power and of the forced cooling flow is determined on the sensor calibration curves. A more sensitive sensor leads to a quadratic calibration curve. This behavior difference of the two calorimetric configurations is explained by means of temperature and heat flux measurements performed with a new instrumented jacket. (authors)

Measuring the linear heat generation rate of a nuclear reactor fuel pin

International Nuclear Information System (INIS)

Smith, R.D.

1981-01-01

A miniature gamma thermometer is described which is capable of travelling through bores distributed in an array through a nuclear reactor core and measure the linear heat generation rate of the fuel pins. (U.K.)

Utilising heat from nuclear waste for space heating

International Nuclear Information System (INIS)

Deacon, D.

1982-01-01

A heating unit utilising the decay heat from irradiated material comprises a storage envelope for the material associated with a heat exchange system, means for producing a flow of air over the heat exchange system to extract heat from the material, an exhaust duct capable of discharging the heated air to the atmosphere, and means for selectively diverting at least some of the heated air to effect the required heating. With the flow of air over the heat exchange system taking place by a natural thermosyphon process the arrangement is self regulating and inherently reliable. (author)

Statement on nuclear incidents at nuclear installations -third quarter 1987

International Nuclear Information System (INIS)

1988-01-01

Four incidents are reported. Carbon dioxide containing a small amount of radioactive material was released at the Chapelcross works of British Nuclear Fuels. However, air monitoring showed the contamination to be small. Surfaces contaminated were cleaned and the reactor was allowed to restart as scheduled. Refurbishment work at the Berkeley Nuclear Power Station sewage plant revealed five radioactive particles. The sludge has been isolated until a disposal method is found. Contamination on a coverall was found at the Sellafield reprocessing plant. Both the operators who may have worn the coverall may have received excess doses. Improvements in the possible source of contamination have been made. A leak of gas from the Hunterston B nuclear power station was found to be from a failed joint on a bolted flange on the bypass circuit heat exchanger. The joint has been repaired. (U.K.)

Magnetic properties of atmospheric PMx in a small settlement during heating and non-heating season

Science.gov (United States)

Petrovsky, E.; Kotlik, B.; Zboril, R.; Kapicka, A.; Grison, H.

2012-04-01

Magnetic properties of environmental samples can serve as fast and relatively cheap proxy method to investigate occurrence of iron oxides. These methods are very sensitive in detecting strongly magnetic compounds such as magnetite and maghemite and can reveal concentration and assess grain-size distribution of these minerals. This information can be significant in estimating e.g. the source of pollutants, monitoring pollution load, or investigating seasonal and climatic effects. We studied magnetic properties of PM1, PM2.5 and PM10, collected over 32-48 hours in a small settlement in south Bohemia during heating and non-heating season. The site is rather remote, with negligible traffic and industrial contributions to air pollution. Thus, the suggested seasonal effect should be dominantly due to local (domestic) heating, burning wood or coal. In our contribution we show typical differences in PMx concentration, which is much higher in the winter (heating) sample, accompanied by SEM analyses and magnetic data oriented on concentration and grain-size distribution of magnetite/maghemite particles. While concentration of Fe-oxides does not vary that much, significant seasonal differences were observed in composition and grain-size distribution, reflecting different sources of the dust particles.

Economics of long-distance transmission, storage, and distribution of heat from nuclear plants with existing and newer techniques

International Nuclear Information System (INIS)

Margen, P.H.

1978-01-01

Conventional and newer types of hot-water pipes are applied to the bulk transport of reject heat from central nuclear power plants to the district heating network of cities or groups of cities. With conventional pipes, the transport of 300 to 2000 MW of heat over distances of 30 to 100 km can be justified, while with newer pipe types, even longer distances would often be economic. For medium-size district heating schemes, low-temperature heat transport from simple heat-only reactors suitable for closer location to cities is of interest. For daily storage of heat on district heating systems, steel heat accumulators are currently used in Sweden. The development of more advanced cheaper heat accumulators, such as lake storage schemes, could make even seasonal heat storage economic. Newer distribution technology extends the economic field of penetration of district heating even to suburban one-family house districts. With proper design and optimization, nuclear district heating can be competitive in a wide market and achieve very substantial fossil-fuel savings

Large-signal, dynamic simulation of the slowpoke-3 nuclear heating reactor

International Nuclear Information System (INIS)

Tseng, C.M.; Lepp, R.M.

1983-07-01

A 2 MWt nuclear reactor, called SLOWPOKE-3, is being developed at the Chalk River Nuclear Laboratories (CRNL). This reactor, which is cooled by natural circulation, is designed to produce hot water for commercial space heating and perhaps generate some electricity in remote locations where the costs of alternate forms of energy are high. A large-signal, dynamic simulation of this reactor, without closed-loop control, was developed and implemented on a hybrid computer, using the basic equations of conservation of mass, energy and momentum. The natural circulation of downcomer flow in the pool was simulated using a special filter, capable of modelling various flow conditions. The simulation was then used to study the intermediate and long-term transient response of SLOWPOKE-3 to large disturbances, such as loss of heat sink, loss of regulation, daily load following, and overcooling of the reactor coolant. Results of the simulation show that none of these disturbances produce hazardous transients

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

Two-phase flow heat transfer in nuclear reactor systems

International Nuclear Information System (INIS)

Koncar, Bostjan; Krepper, Eckhard; Bestion, Dominique; Song, Chul-Hwa; Hassan, Yassin A.

2013-01-01

Complete text of publication follows: Heat transfer and phase change phenomena in two-phase flows are often encountered in nuclear reactor systems and are therefore of paramount importance for their optimal design and safe operation.The complex phenomena observed especially during transient operation of nuclear reactor systems necessitate extensive theoretical and experimental investigations. This special issue brings seven research articles of high quality. Though small in number, they cover a wide range of topics, presenting high complexity and diversity of heat transfer phenomena in two-phase flow. In the last decades a vast amount of research has been devoted to theoretical work and computational simulations, yet the experimental work remains indispensable for understanding of two-phase flow phenomena and for model validation purposes. This is reflected also in this issue, where only one article is purely experimental, while three of them deal with theoretical modelling and the remaining three with numerical simulations. The experimental investigation of the critical heat flux (CHF) phenomena by means of photographic study is presented in the paper of J. Park et al. They have used a high-speed camera system to observe the transient boiling characteristics on a thin horizontal cylinder submerged in a pool of water or highly wetting liquid. Experiments show that the initial boiling process is strongly affected by the properties and wettability of the liquid. The authors have stressed the importance of the local scale observation leading to better understanding of the transient CHF phenomena. In the article of G. Espinosa-Paredes et al. a theoretical work concerning the derivation of transport equations for two-phase flow is presented. The author proposes a novel approach based on derivation of nonlocal volume averaged equations which contain new terms related to nonlocal transport effects. These non-local terms act as coupling elements between the phenomena

Financial analysis of large versus small nuclear power plants

International Nuclear Information System (INIS)

Louh, R.F.; Becker, M.; Wicks, F.

1986-01-01

There have been no new orders for nuclear plants and many nuclear plants under construction have been cancelled in recent years in the United States. Financing problems have been a major factor in this slow down of new nuclear plant activity. Meanwhile, the nuclear plants that have been completed have been operating cost effectively and yielding fossil fuel conservation and air quality benefits. Smaller plants have been designed in the past for the purpose of penetrating markets in developing countries and countries with relatively small utility systems. This paper examines the question of whether these smaller plants would be a viable option to large nuclear plants in the United States. Although the smaller plants are estimated to have a somewhat higher capital cost on a $/k W basis, they have the potential advantage of a lower total financial committment. The computational tools required for this evaluation are optimal generation planning and financial simulation programs and the corresponding generation and financial data bases for a variety of systems

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

Extending nuclear energy to non-electrical applications

Energy Technology Data Exchange (ETDEWEB)

Ingersoll, D.; Houghton, Z. [NuScale Power, LLC, Corvallis, Oregon (United States); Bromm, R. [Fluor Corp., Greenville, SC (United States); Desportes, C. [Aquatech International, Canonsburg, PA (United States); McKellar, M.; Boardman, R. [Idaho National Laboratory, Idaho Falls, ID (United States)

2014-07-01

Electricity represents less than half of all energy consumed in the United States and globally. Although a few commercial nuclear power plants world-wide provide energy to non-electrical applications such as district heating and water desalination, nuclear energy has been largely relegated to base-load electricity production. A new generation of smaller-sized nuclear power plants offers significant promise for extending nuclear energy to many non-electrical applications. The NuScale small modular reactor design is especially well suited for these nontraditional customers due to its small unit size, very robust reactor protection features and a highly flexible and scalable plant design. A series of technical and economic evaluation studies have been conducted to assess the practicality of using a NuScale plant to provide electricity and heat to a variety of non-electrical applications, including water desalination, oil refining, and hydrogen production. The studies serve to highlight the unique design features of the NuScale plant for these applications and provide encouraging conclusions regarding the technical and economic viability of extending clean nuclear energy to a broad range of non-electrical energy consumers. (author)

EXTENDING NUCLEAR ENERGY TO NON-ELECTRICAL APPLICATIONS

Energy Technology Data Exchange (ETDEWEB)

R. Boardman; M. McKellar; D. Ingersoll; Z. Houghton; , R. Bromm; C. Desportes

2014-09-01

Electricity represents less than half of all energy consumed in the United States and globally. Although a few commercial nuclear power plants world-wide provide energy to non-electrical applications such as district heating and water desalination, nuclear energy has been largely relegated to base-load electricity production. A new generation of smaller-sized nuclear power plants offers significant promise for extending nuclear energy to many non-electrical applications. The NuScale small modular reactor design is especially well suited for these non-traditional customers due to its small unit size, very robust reactor protection features and a highly flexible and scalable plant design. A series of technical and economic evaluation studies have been conducted to assess the practicality of using a NuScale plant to provide electricity and heat to a variety of non-electrical applications, including water desalination, oil refining, and hydrogen production. The studies serve to highlight the unique design features of the NuScale plant for these applications and provide encouraging conclusions regarding the technical and economic viability of extending clean nuclear energy to a broad range of non-electrical energy consumers.

Implementation of the project for the construction and operation of a nuclear heat and power plant on the basis of a floating power unit with KLT-40C reactors

Energy Technology Data Exchange (ETDEWEB)

Polushkin, A K; Kuzin, E A [JSC Malaya Energetika, Moscow (Russian Federation); Vorobiov, V M [JSC Atomenergo, St. Petersburg (Russian Federation); Klykov, D M [JSC Iceberg, St. Petersburg (Russian Federation); Panov, J K [OKBM, Nizhny Novgorod (Russian Federation)

2000-09-01

This paper presents the results of research and development on floating nuclear power plant (FNPP) for electricity and heat production for remote locations and small island or coastal communities. Evaluations of construction period, social and economic factors as well as safety and operational issues of the non-self-propelled barge-mounted NPP is given. (author)

Implementation of the project for the construction and operation of a nuclear heat and power plant on the basis of a floating power unit with KLT-40C reactors

International Nuclear Information System (INIS)

Polushkin, A.K.; Kuzin, E.A.; Vorobiov, V.M.; Klykov, D.M.; Panov, J.K.

2000-01-01

This paper presents the results of research and development on floating nuclear power plant (FNPP) for electricity and heat production for remote locations and small island or coastal communities. Evaluations of construction period, social and economic factors as well as safety and operational issues of the non-self-propelled barge-mounted NPP is given. (author)

Heat dissipating nuclear reactor with metal liner

Science.gov (United States)

Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

1985-11-21

A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

In-core program for on line measurements of neutron, photon and nuclear heating parameters inside Jules Horowitz MTR reactor

International Nuclear Information System (INIS)

Lyoussi, A.; Reynard-Carette, C.

2014-01-01

Accurate on-line measurements of key parameters inside experimental channels of Material Testing Reactor are necessary to dimension the irradiation devices and consequently to conduct smart experiments on fuels and materials under suitable conditions. In particular the quantification of nuclear heating, a relevant parameter to reach adapted thermal conditions, has to be improved. These works focus on an important collaborative program between CEA and Aix-Marseille University called INCORE (Instrumentation for Nuclear radiations and Calorimetry On-line in Reactor) dedicated to the development of a new measurement methodology to quantify both nuclear heating and accurate radiation flux levels (neutrons and photons). The methodology, which is based on experiments carried out under irradiation conditions with a multi-sensor device (ionization chamber, fission chamber, gamma thermometer, calorimeter, SPND, SPGD) as well as works performed out-of nuclear/radiative environment on a reference sensor used to measure nuclear heating (calorimeter), is presented (authors)

Industrial space heating and cooling from stored spent nuclear power plant fuel

International Nuclear Information System (INIS)

Shaver, B.O.; Doman, J.W.

1980-01-01

Projections by the Department of Energy indicate that some 5800 metric tons of spent fuel from nuclear power reactors are now in storage and that some 33000 metric tons are expected to be in storage in 1990. The bulk of the spent fuel is currently stored in water-filled basins at the reactor sites from which the material was discharged. The thermal energy in the fuel is dissipated to atmospheres via a pumped water-to-air heat exchanger system. This paper describes a feasibility study of potential methods for the use of the heat. Also, potential applications of heat recovery systems at larger AFR storage facilities were investigated

Development of a control logic for nuclear heating operation for primary system for SMART

International Nuclear Information System (INIS)

Seo, Jae Kwang; Kang, H. O.; Yoon, J. H.; Kim, K. K.; Lee, D. J.

2000-11-01

A nuclear heating concept is adopted in the SMART compared with the commercial nuclear power plant using the primary coolant pumps for heating the primary system. In this report, five options of heatup control logic are proposed and each option is evaluated using MMS code. In option 1, control rod is controlled by a signal of difference in require heatup rate (dT/dt)req. and actual heatup rate (dT/dt)act., which is calculated from the measured value of core outlet temperature. In option 2, control rod is controlled by a signal of difference in reference temperature and actual measured temperature. In option 3, control rod is controlled by a signal of difference in required core power Qcore and actual measured core power N. Primary side temperature difference in measured values between steam generator (SG) inlet and outlet is used in determining Qcore in option 3. Because of this dependency on difference in measured temperature Qcore, in conjunction with measurement channel error in temperature, involves certain uncertainty during specially low flow conditions where primary side temperature difference in SG inlet and outlet is very small. Option 4 is a modified version of option 3. In option 4, SG outlet temperature is not needed to calculate Qcore. However a compensating program which enable Qcore to be evaluated without SG outlet temperature is needed. In option 5, control rod is controlled by a signal of difference in required preset step core power Qcore and actual measured core power N. From the simulation results it is concluded that option 5 using step power setting during heatup operation is suitable for as a heatup control logic for SMART

Analysis of near optimum design for small and medium size nuclear power plants

International Nuclear Information System (INIS)

Ahmed, A.A.

1977-01-01

Market surveys in recent years have shown that a significant market would exist among the developing nations of the world for nuclear power plants that would be classified as small to medium sized, provided that these small plants could produce electricity at a unit price comparable to that of equivalent sized fossil fired plants. Nuclear plants in the range of 100 MWe to 500 MWe would fit more effectively into the relatively smaller grids of most developing nations than would the 900 MWe to 1300 MWe units now being constructed in the large industrial nations. Worldwide re-evaluation of the worth of fossil fuels has prompted a re-examination of the competitive position of small to medium sized nuclear generating units compared to comparable fossil fired units, especially in the context of units specifically optimized for the size range of interest, rather than of designs that are simply scaled down versions of the currently available larger units. Since the absolute cost of electricity is more sensitive to external factors such as cost of money, national inflation rate and time required for licensing and construction than to details of design or perhaps even to choice of fuels, and since the cost of electricity generated in small to medium sized fossil fired units is periodically compared to that of scaled down versions of conventional large nuclear units, the point of view taken here is one of comparing the relative generating costs of smaller nuclear units of optimum design with the corresponding costs of scaled down versions of current large nuclear generating units

Siting analysis and risk assessment for small single-purpose heating reactors

International Nuclear Information System (INIS)

Tarjanne, R.

1979-04-01

Two alternative sites both 10km away from the centre of Helsinki are considered for reactor unit sizes of 400mw and 800mw. The risks associated with a small single-purpose heating reactor is evaluated for normal operation and accident conditions. The evaluation for accident condition is performed for three characteristics accidents. Three pathways are considered in the calculation of the radiation exposure: direct external gamma dose from the release plume, direct gamma radiation from deposited activity on the ground and dose due to inhalation. The risks are compared with the risks from alternative conventional fossil fuelled district heat production methods. The results show that the heating reactor alternative causes an unsignificant risk, which is far less than the risk caused by the fossil-fuelled alternatives

3D Numerical study of the external flow effect on the heat transfer in a radiometric calorimeter dedicated to nuclear heating measurements

International Nuclear Information System (INIS)

Muraglia, M.; Reynard-Carette, C.; Brun, J.; Carette, M.; Lyoussi, A.

2013-06-01

Improvement of measurements in reactor is still a challenge. Thus, this work focuses on numerical studies of one sensor dedicated to nuclear heating measurements: a radiometric complex calorimeter. More precisely, using a simplified conduction heat model, this work presents the first full 3D simulations of a simplified calorimeter reduced to the complex calorimeter head showing that the key parameter for the sensitivity control is the convective heat transfers between the calorimeter and its external surrounding. The effect of external flow velocity on the calorimeter head response is determined for different flow regimes (natural convection, forced convection) and numerical results are found to be in agreement with experimental results under non-irradiated conditions obtained for the complex calorimeter. Moreover, in order to understand and describe fully the mechanisms leading at the different calorimeter heat transfer, the flow velocity dynamics should be added in the model. In a first approach, due to low influence of the flow velocity for tested power range, a static cooling fluid around the calorimeter head is added in the model. Then, in order to get the full flow dynamics, using Boussinesq approximation, a new 2D fluid model, including both temperature field and flow velocity dynamics, is derived taking into account the nuclear heating effect on the flow. (authors)

Potential industrial market for process heat from nuclear reactors

International Nuclear Information System (INIS)

Barnes, R.W.

1976-07-01

A specific segment of industrial process heat use has been examined in detail to identify individual plant locations throughout the United states where nuclear generated steam may be a viable alternative. Five major industries have been studied: paper, chemicals, petroleum, rubber, and primary metals. For these industries, representing 75 percent of the total industrial steam consumption, the individual plant locations within the U.S. using steam in large quantities have been located and characterized as to fuel requirements

Progress and safety aspects in process heat utilization from nuclear systems

International Nuclear Information System (INIS)

Barnert, H.

1995-01-01

Report about the Status and the Progress in the Various Programs and Projects in the Federal Republic of Germany in Process Heat Utilization from the High Temperature Reactor and on Recent Changes of the Atomic Law in the Federal Republic of Germany with Big Influence on the Safety of Nuclear Energy Technology. (author)

Order in large and chaos in small components of nuclear wave functions

International Nuclear Information System (INIS)

Soloviev, V.G.

1992-06-01

An investigation of the order and chaos of the nuclear excited states has shown that there is order in the large and chaos in the small quasiparticle or phonon components of the nuclear wave functions. The order-to-chaos transition is treated as a transition from the large to the small components of the nuclear wave function. The analysis has shown that relatively large many-quasiparticle components of the wave function at an excitation energy (4-8)MeV may exist. The large many-quasiparticle components of the wave functions of the neutron resonances are responsible for enhanced E1-, M1- and E2-transition probabilities from neutron resonance to levels lying (1-2)MeV below them. (author)

On usage of heat-condensation type nuclear heat-and-power plants with the TK type turbines

International Nuclear Information System (INIS)

Boldyrev, V.M.; Smirnov, I.A.; Fedyaev, A.V.; Khrilev, L.S.

1978-01-01

The problems of the efficiency of nuclear heat-and-plants (NHPP) in the heat-andpower energetics of the USSR are discussed. Most attention is centered on an NHPP of heat-condensation type equipped with constant steam flow turbines of the TK-450/500-60 and K-500-60 types and WWER-1000 reactors. According to the specially developed procedure, the problem of selecting the profile of a TK-type turbine, NHPP composition and applications are subjected to the technico-economic analysis. The distance to the urban area from a central heat-and-power plant utilizing organic and atomic fuel is adopted to be the same and equal to 5, 10 and 15 km, and the thermal load is variable between 500 and 7000 Gcal/hour (the share of hot water supply load in the total thermal load being 0.2). The heat supply system is open-circuited, the hot/return water temperatures being 150/70 deg C. The optimum calculated heat production factor for the NHPP does not exceed 0.7, and the optimum heat production values from controlled turbine outputs are within 500-600 Gcal/hour. The mininum thermal load, for which the NHPP with TK turbines is more effective than an organic fuel heat-and-power station, is about 1000-1500 Gcal/hour if cooling towers are used in the technical water supply system, and if it is possible to construct a water reservoir-cooler for the NHPP, this range is reduced to a thermal load level, at which the combined system becomes more effective than the separate power generation systems, i.e. to 500-600 Gcal/hour

Considerations about the utilization of electrically heated rods used for simulation of nuclear fuel pins

International Nuclear Information System (INIS)

Lima, R. de C.F. de; Carajilescov, P.

1987-01-01

The dinamic behavior of electrically heated rods used for simulation of nuclear fuel pins in nuclear power transients, is analysed by the application of the lumped parameter and the finite difference methods. Deviations of the rods surface conditions, for extreme accidental transient conditions are presented and discussed. (author) [pt

Modeling of the heat transfer performance of plate-type dispersion nuclear fuel elements

Science.gov (United States)

Ding, Shurong; Huo, Yongzhong; Yan, XiaoQing

2009-08-01

Considering the mutual actions between fuel particles and the metal matrix, the three-dimensional finite element models are developed to simulate the heat transfer behaviors of dispersion nuclear fuel plates. The research results indicate that the temperatures of the fuel plate might rise more distinctly with considering the particle swelling and the degraded surface heat transfer coefficients with increasing burnup; the local heating phenomenon within the particles appears when their thermal conductivities are too low. With rise of the surface heat transfer coefficients, the temperatures within the fuel plate decrease; the temperatures of the fuel plate are sensitive to the variations of the heat transfer coefficients whose values are lower, but their effects are weakened and slight when the heat transfer coefficients increase and reach a certain extent. Increasing the heat generation rate leads to elevating the internal temperatures. The temperatures and the maximum temperature differences within the plate increase along with the particle volume fractions. The surface thermal flux goes up along with particle volume fractions and heat generation rates, but the effects of surface heat transfer coefficients are not evident.

Proceedings of the International Conference: Nuclear option in countries with small and medium electricity grid

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-12-31

The conference of Croatian Nuclear Society ``Nuclear option in countries with small and medium electricity grid`` is based on experience from last conference of Croatian Nuclear Society in Opatija and on the same philosophy of serving the needs of small or medium present or future user countries. Session topics reflect some current emphasis, such as accommodation of Kyoto restriction on CO{sub 2} emission, or liability and insurance for nuclear damage. In order to achieve best safety and operational standards these countries with limited human and material resources must put greater emphasis on their rational and efficient use. Consequently the world wide developments on innovative reactors` systems and improved concepts for fuel utilisation and waste disposal are substantial interest. Appropriate selections of reactor technology, fuel cycle and decommission strategies are of paramount importance. There are very successful examples of achieving safety and good operational records, so the exchange of experience and cooperation amongst that group of countries would be of great value. As in the future of nuclear energy there will be many more countries with only small or medium nuclear systems, collecting specific experience and cooperation between the like countries will be an additional value to the now prevailing equipment supplier - national utility relationships. Here is presented nine sessions: 1. Energy Options in Countries with Small and Medium Grids 2. Reactors for Small and Medium Electricity Grids 3. Operation and Maintenance Experience 4. Deterministic Safety Analysis 5. Probabilistic Safety Analysis 6. Radioactive Waste Management and Decommissioning 7. Public Relations 8. Emergency Preparedness 9. Liability and Insurance for Nuclear Damage

Proceedings of the International Conference: Nuclear option in countries with small and medium electricity grid

International Nuclear Information System (INIS)

1998-01-01

The conference of Croatian Nuclear Society ''Nuclear option in countries with small and medium electricity grid'' is based on experience from last conference of Croatian Nuclear Society in Opatija and on the same philosophy of serving the needs of small or medium present or future user countries. Session topics reflect some current emphasis, such as accommodation of Kyoto restriction on CO 2 emission, or liability and insurance for nuclear damage. In order to achieve best safety and operational standards these countries with limited human and material resources must put greater emphasis on their rational and efficient use. Consequently the world wide developments on innovative reactors' systems and improved concepts for fuel utilisation and waste disposal are substantial interest. Appropriate selections of reactor technology, fuel cycle and decommission strategies are of paramount importance. There are very successful examples of achieving safety and good operational records, so the exchange of experience and cooperation amongst that group of countries would be of great value. As in the future of nuclear energy there will be many more countries with only small or medium nuclear systems, collecting specific experience and cooperation between the like countries will be an additional value to the now prevailing equipment supplier - national utility relationships. Here is presented nine sessions: 1. Energy Options in Countries with Small and Medium Grids 2. Reactors for Small and Medium Electricity Grids 3. Operation and Maintenance Experience 4. Deterministic Safety Analysis 5. Probabilistic Safety Analysis 6. Radioactive Waste Management and Decommissioning 7. Public Relations 8. Emergency Preparedness 9. Liability and Insurance for Nuclear Damage

New developments in two-phase flow heat transfer with emphasis on nuclear safety research

International Nuclear Information System (INIS)

Mayinger, F.

1987-01-01

The literature on two-phase flow - with and without heat transfer - shows an explosive-like growth of published papers within the last ten years. Many of these papers were published as a result of nuclear safety research. It is impossible to deal with all new developments reported in this extensive literature. So one has to ask: Are there trends of special interest, where this report could be concentrated on? Looking over the situation, there seem to be three very promising fields of research having high actuality, especially for nuclear safety, namely: fluiddynamic and thermodynamic nonequilibrium in steady state, transient conditions, and scaling. The discussion on new developments in two-phase flow heat transfer, therefore, is limited on these subjects

A dilution refrigerator combining low base temperature, high cooling power and low heat leak for use with nuclear cooling

International Nuclear Information System (INIS)

Bradley, D.I.; Guenault, A.M.; Keith, V.; Miller, I.E.; Pickett, G.R.; Bradshaw, T.W.; Locke-Scobie, B.G.

1982-01-01

The design philosophy, design, construction and performance of a dilution refrigerator specifically intended for nuclear cooling experiments in the submillikelvin regime is described. Attention has been paid from the outset to minimizing sources of heat leaks, and to achieving a low base temperature and relatively high cooling power below 10 mK. The refrigerator uses sintered silver heat exchangers similar to those developed at Grenoble. The machine has a base temperature of 3 mK or lower and can precool a copper nuclear specimen in 6.8 T to 8 mK in 70 h. The heat leak to the innermost nuclear stage is < 30 pW after only a few days' running. (author)

Preliminary design analysis of hot gas ducts and a intermediate heat exchanger for the nuclear hydrogen reactor

International Nuclear Information System (INIS)

Song, K. N.; Kim, Y. W.

2008-01-01

Korea Atomic Energy Research Institute (KAERI) is in the process of carrying out a nuclear hydrogen system by considering the indirect cycle gas cooled reactors that produce heat at temperatures in the order of 950 .deg. C. Primary and secondary hot gas ducts with coaxial double tubes and are key components connecting a reactor pressure vessel and a intermediate heat exchanger for the nuclear hydrogen system. In this study, preliminary design analyses on the hot gas ducts and the intermediate heat exchanger were carried out. These preliminary design activities include a preliminary design on the geometric dimensions, a preliminary strength evaluation, thermal sizing, and an appropriate material selection

Manning designs for nuclear district-heating plant (NDHP) with RUTA-type reactor

International Nuclear Information System (INIS)

Gerasimova, V.S.; Mikhan, V.I.; Romenkov, A.A.

2001-01-01

RUTA-type reactor is a water cooled water-moderated pool-type reactor with an atmospheric pressure air medium. The reactor has been designed for heating and hot water supply. Nuclear district heating plant (NDHP) with RUTA-type reactor facility has been designed with a three circuit layout. Primary circuit components are arranged integrally in the reactor vessel. Natural coolant circulation mode is used in the primary circuit. A peculiarity of RUTA-based NDHP as engineered system is a smooth nature of its running slow variation of the parameters at transients. Necessary automation with application of computer equipment will be provided for control and monitoring of heat production process at NDHP. Under developing RUTA-based NDHP it is foreseen that operating staff performs control and monitoring of heat generation process and heat output to consumers as well as current maintenance of NDHP components. All other works associated with NDHP operation should be fulfilled by extraneous personnel. In so doing the participation of operating staff is also possible. (author)

Simulation and Comparison of the Calorimeters Measuring the Nuclear Heating in the OSIRIS Reactor, with the TRIPOLI-4R Monte-Carlo Code

International Nuclear Information System (INIS)

Peron, A.; Malouch, F.; Diop, C.M.

2013-06-01

Two calorimeter devices are used in the OSIRIS MTR reactor (CEA-Saclay center) for the nuclear heating measurements. The first one is a fixed five-stage calorimeter device. The second one is an innovative mobile probe called 'CALMOS'. The design of these devices is different (in particular their geometry), implying modifications on the local neutron and photon fluxes and hence on nuclear heating measured values. The measurements performed by the two calorimeter devices cannot directly be compared; this requires perfect irradiation conditions in the reactor core, especially for the core loading and the control element positions. Simulation is here a good help to perform a fully relevant comparison. In this paper, differences between calorimeter devices in terms of nuclear heating and particle fluxes are evaluated using the TRIPOLI-4 Monte-Carlo code. After a description of the OSIRIS reactor and the design of the two calorimeter devices, the nuclear heating calculation scheme used for simulation will be introduced. Different simulations and results will be detailed and analyzed to determine the calorimeter geometry impact on the measured nuclear heating. (authors)

Prediction of heat and mass transfer in innovative nuclear reactors

International Nuclear Information System (INIS)

Ambrosini, W.; Forgione, N.; Manfredini, A.; Oriolo, F.

2000-01-01

This paper proposes a short review of the different forms adopted to express the analogy between heat and mass transfer for application in correlating data from condensation and evaporation experiments. In particular, the assumptions at the basis of the various forms presented by classical textbooks as well as recent research work are qualitatively discussed, proposing a unified treatment of the different models. On this background, the results of the application of one of the considered forms of the analogy to a problem having relevance for nuclear reactor safety are then discussed. The work performed in this frame is related to condensation on finned tube heat exchangers, proposed as key components in passive containment cooling systems adopted in some innovative reactor concepts. The application of the model to the experimental dana also allowed to obtain interesting information about the effect of different parameters on the cooling capabilities of this compact heat exchangers. (author)

Certification of materials for steam generator condensor and regeneration heat exchanger for nuclear plant

International Nuclear Information System (INIS)

Stevanovicj, M.V.; Jovashevicj, V.J.; Jovashevicj, V.D.J.; Spasicj, Zh.Lj.

1977-01-01

In the construction of a nuclear power plant almost all known materials are used. The choice depends on working conditions. In this work standard specifications of contemporary materials that take part in larger quantities in the following components of the secondary circuit of PWR-type nuclear power plant are proposed: steam generator with moisture separator, condensor and regenerative heat eXchanger

Exergy analysis of a system using a chemical heat pump to link a supercritical water-cooled nuclear reactor and a thermochemical water splitting cycle

International Nuclear Information System (INIS)

Granovskii, M.; Dincer, I.; Rosen, M. A.; Pioro, I

2007-01-01

The power generation efficiency of nuclear plants is mainly determined by the permissible temperatures and pressures of the nuclear reactor fuel and coolants. These parameters are limited by materials properties and corrosion rates and their effect on nuclear reactor safety. The advanced materials for the next generation of CANDU reactors, which employ steam as a coolant and heat carrier, permit the increased steam parameters (outlet temperature up to 625 degree C and pressure of about 25 MPa). Supercritical water-cooled (SCW) nuclear power plants are expected to increase the power generation efficiency from 35 to 45%. Supercritical water-cooled nuclear reactors can be linked to thermochemical water splitting cycles for hydrogen production. An increased steam temperature from the nuclear reactor makes it also possible to utilize its energy in thermochemical water splitting cycles. These cycles are considered by many as one of the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require a heat supply at the temperatures over 550-600 degree C. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump which increases the temperature the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. A high temperature chemical heat pump which employs the reversible catalytic methane conversion reaction is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with a SCW nuclear plant on one side and thermochemical water splitting cycle on the other, increases the temperature level of the 'nuclear' heat and, thus, the intensity of

A Personal Computer-Based Simulator for Nuclear-Heating Reactors