Study of advanced fission power reactor development for the United States. Volume I

International Nuclear Information System (INIS)

1976-01-01

This volume summarizes the results and conclusions of an assessment of five advanced fission power reactor concepts in the context of potential nuclear power economies developed over the time period 1975 to 2020. The study was based on the premise that the LMFBR program has been determined to be the highest priority fission reactor program and it will proceed essentially as planned. Accepting this fact, the overall objective of the study was to provide evaluations of advanced fission reactor systems for input to evaluating the levels of research and development funding for fission power. Evaluation of the reactor systems included the following categories: (1) power plant performance, (2) fuel resource utilization; (3) fuel-cycle requirements; (4) economics; (5) environmental impact; (6) risk to the public; and (7) R and D requirements to achieve commercial status. The specific major objectives of the study were twofold: (1) to parametrically assess the impact of various reactor types for various levels of power demand through the year 2020 on fissile fuel utilization, economics, and the environment, based on varying but reasonable assumptions on the rates of installation; and (2) to qualitatively assess the practicality of the advanced reactor concepts, and their research and development. The reactor concepts examined were limited to the following: advanced high-temperature, gas-cooled reactor (HTGR) systems including the thorium/U-233 fuel cycle, gas turbine, and binary cycle (BIHTGR); gas-cooled fast breeder reactor (GCFR); molten salt breeder reactor (MSBR); light water breeder reactor (LWBR); and CANDU heavy water reactor

Reference reactor module for NASA's lunar surface fission power system

International Nuclear Information System (INIS)

Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Werner, James; Qualls, Louis; Radel, Ross

2009-01-01

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO 2 -fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

Material challenges for the next generation of fission reactor systems

International Nuclear Information System (INIS)

Buckthorpe, Derek

2010-01-01

The new generation of fission reactor systems wil require the deployment and construction of a series of advanced water cooled reactors as part of a package of measures to meet UK and European energy needs and to provide a near term non-fossil fuel power solution that addresses CO 2 emission limits. In addition new longer term Generation IV reactor tye systems are being developed and evaluated to enhance safety, reliability, sustainability economics and proliferation resistance requirements and to meet alternative energy applications (outside of electricity generation) such as process heat and large scale hydrogen generation. New fission systems will impose significant challenges on materials supply and development. In the near term, because of the need to 'gear up' to large scale construction after decades of industrial hibernation/contraction and, in the longer term, because of the need for materials to operate under more challenging environments requiring the deployment and development of new alternative materials not yet established to an industrial stage. This paper investigates the materials challenges imposed by the new Generation III+ and Generation IV systems. These include supply and fabrication issues, development of new high temperature alloys and non-metallic materials, the use of new methods of manufacture and the best use of currently available resources and minerals. Recommendations are made as to how these materials challenges might be met and how governments, industry, manufacturers and researchers can all play their part. (orig.)

Material challenges for the next generation of fission reactor systems

Energy Technology Data Exchange (ETDEWEB)

Buckthorpe, Derek [AMEC, Knutsford, Cheshire (United Kingdom)

2010-07-01

The new generation of fission reactor systems wil require the deployment and construction of a series of advanced water cooled reactors as part of a package of measures to meet UK and European energy needs and to provide a near term non-fossil fuel power solution that addresses CO{sub 2} emission limits. In addition new longer term Generation IV reactor tye systems are being developed and evaluated to enhance safety, reliability, sustainability economics and proliferation resistance requirements and to meet alternative energy applications (outside of electricity generation) such as process heat and large scale hydrogen generation. New fission systems will impose significant challenges on materials supply and development. In the near term, because of the need to 'gear up' to large scale construction after decades of industrial hibernation/contraction and, in the longer term, because of the need for materials to operate under more challenging environments requiring the deployment and development of new alternative materials not yet established to an industrial stage. This paper investigates the materials challenges imposed by the new Generation III+ and Generation IV systems. These include supply and fabrication issues, development of new high temperature alloys and non-metallic materials, the use of new methods of manufacture and the best use of currently available resources and minerals. Recommendations are made as to how these materials challenges might be met and how governments, industry, manufacturers and researchers can all play their part. (orig.)

An optimized symbiotic fusion and molten-salt fission reactor system

International Nuclear Information System (INIS)

Blinkin, V.L.; Novikov, V.M.

A symbiotic fusion-fission reactor system which breeds nuclear fuel is discussed. In the blanket of the controlled thermonuclear reactor (CTR) uranium-233 is generated from thorium, which circulates in the form of ThF 4 mixed with molten sodium and beryllium fluorides. The molten-salt fission reactor (MSR) burns up the uranium-233 and generates tritium for the fusion reactor from lithium, which circulates in the form of LiF mixed with BeF 2 and 233 UF 4 through the MSR core. With a CTR-MSR thermal power ratio of 1:11 the system can produce electrical energy and breed fuel with a doubling time of 4-5 years. The system has the following special features: (1) Fuel reprocessing is much simpler and cheaper than for contemporary fission reactors; reprocessing consists simply in continuous removal of 233 U from the salt circulating in the CTR blanket by the fluorination method and removal of xenon from the MSR fuel salt by gas scavenging; the MSR fuel salt is periodically exchanged for fresh salt and the 233 U is then removed from it; (2) Tritium is produced in the fission reactor, which is a much simpler system than the fusion reactor; (3) The CTR blanket is almost ''clean''; no tritium is produced in it and fission fragment activity does not exceed the activity induced in the structural materials; (4) Almost all the thorium introduced into the CTR blanket can be used for producing 233 U

Some methods of failed fuel element detection in water cooled reactors

International Nuclear Information System (INIS)

Strindehag, O.M.

1976-01-01

The methods are surveyed using fission products released in the coolant for the detection of failed fuel elements in water cooled reactors. The classification of the detection methods is made with respect to fission product detection in the coolant and to gaseous fission product detection. The detection systems are listed used for the AGESTA power reactor and for the experimental loops of the RA research reactor based on the detection of either gaseous fission products or gaseous daughter products. The AGESTA reactor detection systems using electrostatic precipitators consist of five precipitator channels of which three are intended for detection and two for localization. A special detection unit was developed for the failed fuel element detection in the R-2 reactor experimental steam loop. Its description is listed. In the reactor pressurized-water loop a Cherenkov counter was used in the detection of fission products. An ion exchange monitor whose application is described was used in the total measurement of the main coolant flow in the AGESTA reactor. (J.P.)

Design of a power conversion system for an indirect cycle, helium cooled pebble bed reactor system

International Nuclear Information System (INIS)

Wang, C.; Ballinger, R.G.; Stahle, P.W.; Demetri, E.; Koronowski, M.

2002-01-01

A design is presented for the turbomachinery for an indirect cycle, closed, helium cooled modular pebble bed reactor system. The design makes use of current technology and will operate with an overall efficiency of 45%. The design uses an intermediate heat exchanger which isolated the reactor cycle from the turbomachinery. This design excludes radioactive fission products from the turbomachinery. This minimizes the probability of an air ingress accident and greatly simplifies maintenance. (author)

Development of failure detection system for gas-cooled reactor

International Nuclear Information System (INIS)

Feirreira, M.P.

1990-01-01

This work presents several kinds of Failure Detection Systems for Fuel Elements, stressing their functional principles and major applications. A comparative study indicates that the method of electrostatic precipitation of the fission gases Kr and Xe is the most efficient for fuel failure detection in gas-cooled reactors. A detailed study of the physical phenomena involved in electrostatic precipitation led to the derivation of an equation for the measured counting rate. The emission of fission products from the fuel and the ion recombination inside the chamber are evaluated. A computer program, developed to simulate the complete operation of the system, relates the counting rate to the concentration of Kr and Xe isotopes. The project of a mock-up is then presented. Finally, the program calculations are compared to experimental data, available from the literature, yielding a close agreement. (author)

Synergies in the design and development of fusion and generation IV fission reactors

International Nuclear Information System (INIS)

Bogusch, E.; Carre, F.; Knebel, J.; Aoto, K.

2007-01-01

Future fusion reactors or systems and Generation IV fission reactors are designed and developed in worldwide programmes mostly involving the same partners to investigate and assess their potential for realisation and contribution to meet the future energy needs beyond 2030. Huge scientific and financial effort is necessary to meet these objectives. First programmes have been launched in Generation IV International Forum (GIF) for fission and in the Broader Approach for fusion reactor system development. Except the physics basis for the energy source, future fusion and fission reactors, in particular those with fast neutron core face similar design issues and development needs. Therefore the call for the identification of synergies became evident. Beyond ITER cooled by water, future fusion reactors or systems will be designed for helium and liquid metal cooling and higher temperatures similar to those proposed for some of the six fission reactor concepts in GIF with their diverse coolants. Beside materials developments which are not discussed in this paper, design and performance of components and systems related to the diverse coolants including lifetime and maintenance aspects might offer significant potentials for synergies. Furthermore, the use of process heat for applications in addition to electricity production as well as their safety approaches might create synergistic design and development programmes. Therefore an early identification of possible synergies in the relevant programmes should be endorsed to minimise the effort for future power plants in terms of investments and resources. In addition to a general overview of a possible synergistic work programme which promotes the interaction between fusion and fission programmes towards an integrated organisation of their design and R and D programmes, some specific remarks will be given for joint design tools, numerical code systems and joint experiments in support of common technologies. (orig.)

Synergies in the design and development of fusion and generation IV fission reactors

International Nuclear Information System (INIS)

Bogusch, E.; Carre, F.; Knebel, J.U.; Aoto, K.

2008-01-01

Future fusion reactor and Generation IV fission reactor systems are designed and developed in worldwide programmes to investigate and assess their potential for realisation and contribution to the future energy needs beyond 2030 mostly involving the same partners. Huge scientific and financial effort is necessary to meet these objectives. First programmes have been launched in Generation IV International Forum (GIF) for fission and in the Broader Approach for fusion reactor system development. Except for the physics basis for the energy source, future fusion and fission reactors, in particular those with fast neutron core, face similar design issues and development needs. Therefore, the call for the identification of synergies became evident. Beyond ITER cooled by water, future fusion reactor systems will be designed for high-temperature helium and liquid metal cooling but also water including supercritical water and molten salt similar to those proposed for some of the six fission reactor concepts in GIF with their diverse coolants. Beside materials developments which are not discussed in this paper, design and performance of components and systems related to the diverse coolants including lifetime and maintenance aspects might offer significant potentials for synergies. Furthermore, the use of process heat for applications in addition to electricity production as well as their safety approaches can create synergistic design and development programmes. Therefore, an early identification of possible synergies in the relevant programmes should be endorsed to minimise the effort for future power plants in terms of investments and resources. In addition to a general overview of a possible synergistic work programme which promotes the interaction between fusion and fission programmes towards an integrated organisation of their design and R and D programmes, some specific remarks will be given for joint design tools, numerical code systems and joint experiments in

Cooling system for auxiliary reactor component

International Nuclear Information System (INIS)

Fujihira, Tomoko.

1991-01-01

A cooling system for auxiliary reactor components comprises three systems, that is, two systems of reactor component cooling water systems (RCCW systems) and a high pressure component cooling water system (HPCCW system). Connecting pipelines having partition valves are intervened each in a cooling water supply pipeline to an emmergency component of each of the RCCW systems, a cooling water return pipeline from the emmergency component of each of the RCCW systems, a cooling water supply pipeline to each of the emmergency components of one of the RCCW system and the HPCCW system and a cooling water return pipeline from each of the emmergency components of one of the RCCW system and the HPCCW system. With such constitution, cooling water can be supplied also to the emmergency components in the stand-by system upon periodical inspection or ISI, thereby enabling to improve the backup performance of the emmergency cooling system. (I.N.)

Materials degradation in fission reactors: Lessons learned of relevance to fusion reactor systems

International Nuclear Information System (INIS)

Was, Gary S.

2007-01-01

The management of materials in power reactor systems has become a critically important activity in assuring the safe, reliable and economical operation of these facilities. Over the years, the commercial nuclear power reactor industry has faced numerous 'surprises' and unexpected occurrences in materials. Mitigation strategies have sometimes solved one problem at the expense of creating another. Other problems have been solved successfully and have motivated the development of techniques to foresee problems before they occur. This paper focuses on three aspects of fission reactor experience that may benefit future fusion systems. The first is identification of parameters and processes that have had a large impact on the behavior of materials in fission systems such as temperature, dose rate, surface condition, gradients, metallurgical variability and effects of the environment. The second is the development of materials performance and failure models to provide a basis for assuring component integrity. Last is the development of proactive materials management programs that identify and pre-empt degradation processes before they can become problems. These aspects of LWR experience along with the growing experience with materials in the more demanding advanced fission reactor systems form the basis for a set of 'lessons learned' to aid in the successful management of materials in fusion reactor systems

Core cooling system for reactor

International Nuclear Information System (INIS)

Kondo, Ryoichi; Amada, Tatsuo.

1976-01-01

Purpose: To improve the function of residual heat dissipation from the reactor core in case of emergency by providing a secondary cooling system flow channel, through which fluid having been subjected to heat exchange with the fluid flowing in a primary cooling system flow channel flows, with a core residual heat removal system in parallel with a main cooling system provided with a steam generator. Constitution: Heat generated in the core during normal reactor operation is transferred from a primary cooling system flow channel to a secondary cooling system flow channel through a main heat exchanger and then transferred through a steam generator to a water-steam system flow channel. In the event if removal of heat from the core by the main cooling system becomes impossible due to such cause as breakage of the duct line of the primary cooling system flow channel or a trouble in a primary cooling system pump, a flow control valve is opened, and steam generator inlet and outlet valves are closed, thus increasing the flow rate in the core residual heat removal system. Thereafter, a blower is started to cause dissipation of the core residual heat from the flow channel of a system for heat dissipation to atmosphere. (Seki, T.)

Fuel Development For Gas-Cooled Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

M. K. Meyer

2006-06-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

Some safety studies for conceptual fusion--fission hybrid reactors. Final report

International Nuclear Information System (INIS)

Kastenberg, W.E.; Okrent, D.

1978-07-01

The objective of this study was to make a preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors. The study and subsequent analysis was largely based upon reference to one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The blanket is a fast-spectrum, uranium carbide, helium cooled, subcritical reactor, optimized for the production of fissile fuel. An attempt was made to generalize the results wherever possible

Reference reactor module for NASA's lunar surface fission power system

Energy Technology Data Exchange (ETDEWEB)

Poston, David I [Los Alamos National Laboratory; Kapernick, Richard J [Los Alamos National Laboratory; Dixon, David D [Los Alamos National Laboratory; Werner, James [INL; Qualls, Louis [ORNL; Radel, Ross [SNL

2009-01-01

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

Emergency reactor cooling systems for the experimental VHTR

International Nuclear Information System (INIS)

Mitake, Susumu; Suzuki, Katsuo; Miyamoto, Yoshiaki; Tamura, Kazuo; Ezaki, Masahiro.

1983-03-01

Performances and design of the panel cooling system which has been proposed to be equipped as an emergency reactor cooling system for the experimental multi purpose very high temperature gas-cooled reactor are explained. Effects of natural circulation flow which would develop in the core and temperature transients of the panel in starting have been precisely investigated. Conditions and procedures for settling accidents with the proposed panel cooling system have been also studied. Based on these studies, it has been shown that the panel cooling system is effective and useful for the emergency reactor cooling of the experimental VHTR. (author)

Comparative energetics of three fusion-fission symbiotic nuclear reactor systems

International Nuclear Information System (INIS)

Gordon, C.W.; Harms, A.A.

1975-01-01

The energetics of three symbiotic fusion-fission nuclear reactor concepts are investigated. The fuel and power balances are considered for various values of systems parameters. The results from this analysis suggest that symbiotic fusion-fission systems are advantageous from the standpoint of economy and resource utilization. (Auth.)

Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

Science.gov (United States)

Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

MEANS FOR SHIELDING AND COOLING REACTORS

Science.gov (United States)

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1959-02-10

Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

Reactor cooling system

International Nuclear Information System (INIS)

Kato, Etsuji.

1979-01-01

Purpose: To eliminate cleaning steps in the pipelines upon reactor shut-down by connecting a filtrating and desalting device to the cooling system to thereby always clean up the water in the pipelines. Constitution: A filtrating and desalting device is connected to the pipelines in the cooling system by way of drain valves and a check valve. Desalted water is taken out from the exit of the filtrating and desalting device and injected to one end of the cooling system pipelines by way of the drain valve and the check valve and then returned by way of another drain valve to the desalting device. Water in the pipelines is thus always desalted and the cleaning step in the pipelines is no more required in the shut-down. (Kawakami, Y.)

Containment atmosphere cooling system for experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

1979-01-01

The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

IAEA high temperature gas-cooled reactor activities

International Nuclear Information System (INIS)

Kendall, J.M.

2000-01-01

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

IAEA high temperature gas cooled reactor activities

International Nuclear Information System (INIS)

Kendall, J.M.

2001-01-01

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

Study on fission blanket fuel cycling of a fusion-fission hybrid energy generation system

International Nuclear Information System (INIS)

Zhou, Z.; Yang, Y.; Xu, H.

2011-01-01

This paper presents a preliminary study on neutron physics characteristics of a light water cooled fission blanket for a new type subcritical fusion-fission hybrid reactor aiming at electric power generation with low technical limits of fission fuel. The major objective is to study the fission fuel cycling performance in the blanket, which may possess significant impacts on the feasibility of the new concept of fusion-fission hybrid reactor with a high energy gain (M) and tritium breeding ratio (TBR). The COUPLE2 code developed by the Institute of Nuclear and New Energy Technology of Tsinghua University is employed to simulate the neutronic behaviour in the blanket. COUPLE2 combines the particle transport code MCNPX with the fuel depletion code ORIGEN2. The code calculation results show that soft neutron spectrum can yield M > 20 while maintaining TBR >1.15 and the conversion ratio of fissile materials CR > 1 in a reasonably long refuelling cycle (>five years). The preliminary results also indicate that it is rather promising to design a high-performance light water cooled fission blanket of fusion-fission hybrid reactor for electric power generation by directly loading natural or depleted uranium if an ITER-scale tokamak fusion neutron source is achievable.

Conceptual design tool development for a Pb-Bi cooled reactor

International Nuclear Information System (INIS)

Lee, K. G.; Chang, S. H.; No, H. C.; Chunm, M. H.

2000-01-01

Conceptual design is generally ill-structured and mysterious problem solving. This leads the experienced experts to be still responsible for the most of synthesis and analysis task, which are not amenable to logical formulations in design problems. Especially because a novel reactor such as a Pb-Bi cooled reactor is going on a conceptual design stage, it will be very meaningful to develop the conceptual design tool. This tool consists of system design module with artificial intelligence, scaling module, and validation module. System design decides the optimal structure and the layout of a Pb-Bi cooled reactor, using design synthesis part and design analysis part. The designed system is scaled to be optimal with desired power level, and then the design basis accidents (Dbase) are analyzed in validation module. Design synthesis part contains the specific data for reactor components and the general data for a Pb-Bi cooled reactor. Design analysis part contains several design constraints for formulation and solution of a design problem. In addition, designer's intention may be externalized through emphasis on design requirements. For the purpose of demonstration, the conceptual design tool is applied to a Pb-Bi cooled reactor with 125 M Wth of power level. The Pb-Bi cooled reactor is a novel reactor concept in which the fission-generated heat is transferred from the primary coolant to the secondary coolant through a reactor vessel wall of a novel design. The Pb-Bi cooled reactor is to deliver 125 M Wth per module for 15 effective full power years without any on-site fuel handling. The conceptual design tool investigated the feasibility of a Pb-Bi cooled reactor. Application of the conceptual design tool will be, in detail, presented in the full paper. (author)

Emergency cooling system for a gas-cooled nuclear reactor

International Nuclear Information System (INIS)

Cook, R.K.; Burylo, P.S.

1975-01-01

The site of the gas-cooled reactor with direct-circuit gas turbine is preferably the sea coast. An emergency cooling system with safety valve and emergency feed-water addition is designed which affects at least a part of the reactor core coolant after leaving the core. The emergency cooling system includes a water emergency cooling circuit with heat exchanger for the core coolant. The safety valve releases water or steam from the emergency coolant circuit when a certain temperature is exceeded; this is, however, replaced by the emergency feed-water. If the gas turbine exhibits a high and low pressure turbine stage, which are flowed through by coolant one behind another, a part of the coolant can be removed in front of each part turbine by two valves and be added to the haet exchanger. (RW/LH) [de

Summary of ORNL high-temperature gas-cooled reactor program

International Nuclear Information System (INIS)

Kasten, P.R.

1981-01-01

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

FFTF fission gas monitor computer system

International Nuclear Information System (INIS)

Hubbard, J.A.

1987-01-01

The Fast Flux Test Facility (FFTF) is a liquid-metal-cooled test reactor located on the Hanford site. A dual computer system has been developed to monitor the reactor cover gas to detect and characterize any fuel or test pin fission gas releases. The system acquires gamma spectra data, identifies isotopes, calculates specific isotope and overall cover gas activity, presents control room alarms and displays, and records and prints data and analysis reports. The fission gas monitor system makes extensive use of commercially available hardware and software, providing a reliable and easily maintained system. The design provides extensive automation of previous manual operations, reducing the need for operator training and minimizing the potential for operator error. The dual nature of the system allows one monitor to be taken out of service for periodic tests or maintenance without interrupting the overall system functions. A built-in calibrated gamma source can be controlled by the computer, allowing the system to provide rapid system self tests and operational performance reports

Radioactivity Monitoring System for TRIGA 2000 Reactor Water Tank with On-Line Gamma Spectrometer

International Nuclear Information System (INIS)

Prasetyo Basuki; Sudjatmi KA

2009-01-01

One of the requirements in radiological safety in the operating condition of research reactor are the absence of radionuclide from fission product released to reactor cooling water and environment. Early detection of fission product that released from fuel element can be done by monitoring radioactivity level on primary cooling water.Reactor cooling water can be used as an important indicator in detecting radioactivity level of material fission product, when the leakage occurs. Therefore, it needs to build a monitoring system for measuring radioactivity level of cooling water directly and simple. The idea of this system is counting radioactivity water flow from reactor tank to the marinelli cube that attached to the HPGe detector on gamma spectrometer. Cooling water from tank aimed on plastic pipe to the marinelli cube. Water flows in gravitational driven to the marinelli cube, with volume flow rate 5.1 liters/minute in the inlet and 2.2 liters/minute in output. (author)

Systems study of tokamak fusion--fission reactors

International Nuclear Information System (INIS)

Tenney, F.H.; Bathke, C.G.; Price, W.G. Jr.; Bohlke, W.H.; Mills, R.G.; Johnson, E.F.; Todd, A.M.M.; Buchanan, C.H.; Gralnick, S.L.

1978-11-01

This publication reports the results of a two to three year effort at a systematic analysis of a wide variety of tokamak-driven fissioning blanket reactors, i.e., fusion--fission hybrids. It addresses the quantitative problems of determining the economically most desirable mix of the two products: electric power and fissionable fuel and shows how the price of electric power can be minimized when subject to a variety of constraints. An attempt has been made to avoid restricting assumptions, and the result is an optimizing algorithm that operates in a six-dimensional parameter space. Comparisons are made on sets of as many as 100,000 distinct machine models, and the principal results of the study have been derived from the examination of several hundred thousand possible reactor configurations

Core of a liquid-cooled nuclear reactor

International Nuclear Information System (INIS)

Wright, J.R.; McFall, A.

1975-01-01

The core of a liquid-cooled nuclear reactor, e.g. of a sodium-cooled fast reactor, is protected in such a way that the recoil wave resulting from loss of coolant in a cooling channel and caused by released gas is limited to a coolant inlet chamber of this cooling channel. The channels essentially consist of the coolant inlet chamber and a fuel chamber - with a fission gas storage plenum - through which the coolant flows. Between the two chambers, a locking device within a tube is provided offering a much larger flow resistance to the backflow of gas or coolant than in flow direction. The locking device may be a hydraulic countertorque control system, e.g. a valvular line. Other locking devices have got radially helical vanes running around an annular flow space. Furthermore, the locking device may consist of a number of needles running parallel to each other and forming a circular grid. Though it can be expanded by the forward flow - the needles are spreading - , it acts as a solid barrier for backflows. (TK) [de

Conceptual design of the blanket and power conversion system for a mirror hybrid fusion-fission reactor. 12-month progress report, July 1, 1975--June 30, 1976

International Nuclear Information System (INIS)

Schultz, K.R.; Baxi, C.B.; Rao, R.

1976-01-01

This report presents the conceptual design and preliminary feasibility assessment for the hybrid blanket and power conversion system of the Mirror Hybrid Fusion-Fission Reactor. Existing gas-cooled fission reactor technology is directly applicable to the Mirror Hybrid Reactor. There are a number of aspects of the present conceptual design that require further design and analysis effort. The blanket and power conversion system operating parameters have not been optimized. The method of supporting the blanket modules and the interface between these modules and the primary loop helium ducting will require further design work. The means of support and containment of the primary loop components must be studied. Nevertheless, in general, the conceptual design appears quite feasible

An alternative solution for heavy liquid metal cooled reactors fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

2014-10-15

Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

System model for analysis of the mirror fusion-fission reactor

International Nuclear Information System (INIS)

Bender, D.J.; Carlson, G.A.

1977-01-01

This report describes a system model for the mirror fusion-fission reactor. In this model we include a reactor description as well as analyses of capital cost and blanket fuel management. In addition, we provide an economic analysis evaluating the cost of producing the two hybrid products, fissile fuel and electricity. We also furnish the results of a limited parametric analysis of the modeled reactor, illustrating the technological and economic implications of varying some important reactor design parameters

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

International Nuclear Information System (INIS)

Conklin, J.C.

1990-04-01

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

Emergency core cooling system in BWR type reactors

International Nuclear Information System (INIS)

Takizawa, Yoji

1981-01-01

Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

Method of injecting cooling water in emergency core cooling system (ECCS) of PWR type reactor

International Nuclear Information System (INIS)

Sobajima, Makoto; Adachi, Michihiro; Tasaka, Kanji; Suzuki, Mitsuhiro.

1979-01-01

Purpose: To provide a cooling water injection method in an ECCS, which can perform effective cooling of the reactor core. Method: In a method of injecting cooling water in an ECCS as a countermeasure against a rupture accident of a pwr type reactor, cooling water in the first pressure storage injection system is injected into the upper plenum of the reactor pressure vessel at a set pressure of from 50 to 90 atg. and a set temperature of from 80 to 200 0 C, cooling water in the second pressure storage injection system is injected into the lower plenum of the reactor pressure vessel at a pressure of from 25 to 60 atg. which is lower than the set pressure and a temperature less than 60 0 C, and further in combination with these procedures, cooling water of less than 60 0 C is injected into a high-temperature side piping, in the high-pressure injection system of upstroke of 100 atg. by means of a pump and the low-pressure injection system of upstroke of 20 atg. also by means of a pump, thereby cooling the reactor core. (Aizawa, K.)

Hefei experimental hybrid fusion-fission reactor conceptual design

International Nuclear Information System (INIS)

Qiu Lijian; Luan Guishi; Xu Qiang

1992-03-01

A new concept of hybrid reactor is introduced. It uses JET-like(Joint European Tokamak) device worked at sub-breakeven conditions, as a source of high energy neutrons to induce a blanket fission of depleted uranium. The solid breeding material and helium cooling technique are also used. It can produce 100 kg of 239 Pu per year by partial fission suppressed. The energy self-sustained of the fusion core is not necessary. Plasma temperature is maintained by external 20 MW ICRF (ion cyclotron resonance frequency) and 10 MW ECRF (electron cyclotron resonance frequency) heating. A steady state plasma current at 1.5 Ma is driven by 10 MW LHCD (lower hybrid current driven). Plasma density will be kept by pellet injection. ICRF can produce a high energy tail in ion distribution function and lead to significant enhancement of D-T reaction rate by 2 ∼ 5 times so that the neutron source strength reaches to the level of 1 x 10 19 n/s. This system is a passive system. It's power density is 10 W/cm 3 and the wall loading is 0.6 W/cm 2 that is the lower limitation of fusion and fission technology. From the calculation of neutrons it could always be in sub-critical and has intrinsic safety. The radiation damage and neutron flux distribution on the first wall are also analyzed. According to the conceptual design the application of this type hybrid reactor earlier is feasible

The behavior of fission products during nuclear rocket reactor tests

International Nuclear Information System (INIS)

Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

1991-01-01

Fission product release from nuclear rocket propulsion reactor fuel is an important consideration for nuclear rocket development and application. Fission product data from the last six reactors of the Rover program are collected in this paper to provide as basis for addressing development and testing issues. Fission product loss from the fuel will depend on fuel composition and reactor design and operating parameters. During ground testing, fission products can be contained downstream of the reactor. The last Rover reactor tested, the Nuclear Furnance, was mated to an effluent clean-up system that was effective in preventing the discharge of fission products into the atmosphere

Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

Transmutation of fission products in reactors and accelerator-driven systems

International Nuclear Information System (INIS)

Janssen, A.J.

1994-01-01

Energy flows and mass flows in several scenarios are considered. Economical and safety aspects of the transmutation scenarios are compared. It is difficult to find a sound motivation for the transmutation of fission products with accelerator-driven systems. If there would be any hesitation in transmuting fission products in nuclear reactors, there would be an even stronger hesitation to use accelerator-driven systems, mainly because of their lower energy efficiency and their poor cost effectiveness. The use of accelerator-driven systems could become a 'meaningful' option only if nuclear energy would be banished completely. (orig./HP)

Passive afterheat removal in the HTGR with the liner cooling system as a heat sink

International Nuclear Information System (INIS)

Rehm, W.; Jahn, W.; Verfondern, K.

1984-09-01

The report deals with the transients of temperature and system pressure and the fission product behaviour in the primary circuit of an HTGR during passive afterheat removal, where the liner cooling system of the PCRV serves as a heat sink. The analysis has been made for the PNP-500-reactor representing nuclear plants with medium thermal power. The investigations show that the liner cooling system is able to control a core heatup. High temperature loads are encountered in the upper core region. In the case of a reactor under pressure the fuel elements and the primary circuit remain intact as the first and second barriers for fission products. In the case of a depressurized primary circuit the liner cooling system also keeps the PCRV at normal operating temperatures. The effects of a core heatup on component damage and release of fission products are thus limited. (orig.) [de

Reactor-core isolation cooling system with dedicated generator

International Nuclear Information System (INIS)

Nazareno, E.V.; Dillmann, C.W.

1992-01-01

This patent describes a nuclear reactor complex. It comprises a dual-phase nuclear reactor; a main turbine for converting phase-conversion energy stored by vapor into mechanical energy for driving a generator; a main generator for converting the mechanical energy into electricity; a fluid reservoir external to the reactor; a reactor core isolation cooling system with several components at least some of which require electrical power. It also comprises an auxiliary pump for pumping fluid from the reservoir into the reactor pressure vessel; an auxiliary turbine for driving the pump; control means for regulating the rotation rate of the auxiliary turbine; cooling means for cooling the control means; and an auxiliary generator coupled to the auxiliary turbine for providing at least a portion of the electrical power required by the components during a blackout condition

Mirror hybrid reactor blanket and power conversion system conceptual design

International Nuclear Information System (INIS)

Schultz, K.R.; Backus, G.A.; Baxi, C.B.; Dee, J.B.; Estrine, E.A.; Rao, R.; Veca, A.R.

1976-01-01

The conceptual design of the blanket and power conversion system for a gas-cooled mirror hybrid fusion-fission reactor is presented. The designs of the fuel, blanket module and power conversion system are based on existing gas-cooled fission reactor technology that has been developed at General Atomic Company. The uranium silicide fuel is contained in Inconel-clad rods and is cooled by helium gas. The fuel is contained in 16 spherical segment modules which surround the fusion plasma. The hot helium is used to raise steam for a conventional steam cycle turbine generator. The details of the method of support for the massive blanket modules and helium ducts remain to be determined. Nevertheless, the conceptual design appears to be technically feasible with existing gas-cooled technology. A preliminary safety analysis shows that with the development of a satisfactory method of primary coolant circuit containment and support, the hybrid reactor could be licensed under existing Nuclear Regulatory Commission regulations

Fuel cycles and advanced core designs for the Gas-Cooled Fast Breeder Reactor

International Nuclear Information System (INIS)

Simon, R.H.; Hamilton, C.J.; Hunter, R.S.

1982-01-01

Studies indicate that a 1200 MW(e) Gas-Cooled Fast Breeder Reactor could achieve compound system doubling times of under ten years when using advanced oxide or carbide fuels. In addition, when thorium is used in the breeding blankets, enough U-233 can be generated in each GCFR to supply several advanced converter reactors with fissionable material and this symbiotic relationship could provide energy for the world for centuries. (author)

Fission-Based Electric Propulsion for Interstellar Precursor Missions

International Nuclear Information System (INIS)

HOUTS, MICHAEL G.; LENARD, ROGER X.; LIPINSKI, RONALD J.; PATTON, BRUCE; POSTON, DAVID; WRIGHT, STEVEN A.

1999-01-01

This paper reviews the technology options for a fission-based electric propulsion system for interstellar precursor missions. To achieve a total ΔV of more than 100 km/s in less than a decade of thrusting with an electric propulsion system of 10,000s Isp requires a specific mass for the power system of less than 35 kg/kWe. Three possible configurations are described: (1) a UZrH-fueled,NaK-cooled reactor with a steam Rankine conversion system,(2) a UN-fueled gas-cooled reactor with a recuperated Brayton conversion system, and (3) a UN-fueled heat pipe-cooled reactor with a recuperated Brayton conversion system. All three of these systems have the potential to meet the specific mass requirements for interstellar precursor missions in the near term. Advanced versions of a fission-based electric propulsion system might travel as much as several light years in 200 years

Emergency reactor cooling device

International Nuclear Information System (INIS)

Arakawa, Ken.

1993-01-01

An emergency nuclear reactor cooling device comprises a water reservoir, emergency core cooling water pipelines having one end connected to a water feeding sparger, fire extinguishing facility pipelines, cooling water pressurizing pumps, a diesel driving machine for driving the pumps and a battery. In a water reservoir, cooling water is stored by an amount required for cooling the reactor upon emergency and for fire extinguishing, and fire extinguishing facility pipelines connecting the water reservoir and the fire extinguishing facility are in communication with the emergency core cooling water pipelines connected to the water feeding sparger by system connection pipelines. Pumps are operated by a diesel power generator to introduce cooling water from the reservoir to the emergency core cooling water pipelines. Then, even in a case where AC electric power source is entirely lost and the emergency core cooling system can not be used, the diesel driving machine is operated using an exclusive battery, thereby enabling to inject cooling water from the water reservoir to a reactor pressure vessel and a reactor container by the diesel drive pump. (N.H.)

Cooling device for reactor container

International Nuclear Information System (INIS)

Arai, Kenji.

1996-01-01

Upon assembling a static container cooling system to an emergency reactor core cooling system using dynamic pumps in a power plant, the present invention provides a cooling device of lowered center of gravity and having a good cooling effect by lowering the position of a cooling water pool of the static container cooling system. Namely, the emergency reactor core cooling system injects water to the inside of a pressure vessel using emergency cooling water stored in a suppression pool as at least one water source upon loss of reactor coolant accident. In addition, a cooling water pool incorporating a heat exchanger is disposed at the circumference of the suppression pool at the outside of the container. A dry well and the heat exchanger are connected by way of steam supply pipes, and the heat exchanger is connected with the suppression pool by way of a gas exhaustion pipe and a condensate returning pipeline. With such a constitution, the position of the heat exchanger is made higher than an ordinary water level of the suppression pool. As a result, the emergency cooling water of the suppression pool water is injected to the pressure vessel by the operation of the reactor cooling pumps upon loss of coolant accident to cool the reactor core. (I.S.)

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

International Nuclear Information System (INIS)

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

1994-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-04-01

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

Qualitative and quantitative characteristics of fission products in spent nuclear fuel from RBMK-type reactor

International Nuclear Information System (INIS)

Adlys, G.; Adliene, D.

2002-01-01

Well-known empirical models or experimental instruments and methods for the estimation of fission product yields do not allow prediction of the behavior and evaluation of the time-dependent qualitative and quantitative characteristics of all fission products in spent nuclear fuel during long-term storage. Several computer codes were developed in different countries to solve this problem. French codes APOLLO1 and PEPIN were used in this work for modeling the characteristics of spent nuclear fuel in the RBMK reactor. The modeling results of qualitative and quantitative characteristics of long-lived fission products for different cooling periods of spent nuclear fuel, including 50-year cooling period, are presented in this paper. The 50-year cooling period conforms to the foreseen time of storage of spent nuclear fuel in CONSTOR and CASTOR casks at the Ignalina NPP. These results correlate well with evaluated quantities for the well-known yields of the nuclides and could be used for the compilation of the database for long-lived fission products in spent nuclear fuel from the RBMK-type reactor. They allow one to predict and to solve effectively safety problems concerning with long-term spent nuclear fuel storage in casks. (author)

Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Dawn Scates

2010-10-01

A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

A gas-cooled reactor surface power system

International Nuclear Information System (INIS)

Lipinski, R.J.; Wright, S.A.; Lenard, R.X.; Harms, G.A.

1999-01-01

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars. copyright 1999 American Institute of Physics

A gas-cooled reactor surface power system

International Nuclear Information System (INIS)

Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

1999-01-01

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars

A Gas-Cooled Reactor Surface Power System

Energy Technology Data Exchange (ETDEWEB)

Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

1998-11-09

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

Passive cooling system for nuclear reactor containment structure

Science.gov (United States)

Gou, Perng-Fei; Wade, Gentry E.

1989-01-01

A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

Gas-cooled reactors

International Nuclear Information System (INIS)

Vakilian, M.

1977-05-01

The present study is the second part of a general survey of Gas Cooled Reactors (GCRs). In this part, the course of development, overall performance and present development status of High Temperature Gas Cooled Reactors (HTCRs) and advances of HTGR systems are reviewed. (author)

Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

Energy Technology Data Exchange (ETDEWEB)

Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

2017-05-15

Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

International Nuclear Information System (INIS)

Laureau, A.; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

2017-01-01

Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

Major features of a mirror fusion--fast fission hybrid reactor

International Nuclear Information System (INIS)

Moir, R.W.; Lee, J.D.; Burleigh, R.J.

1974-01-01

A conceptual design was made of a fusion-fission reactor. The fusion component is a D-T plasma confined by a pair of magnetic mirror coils in a Yin-Yang configuration and sustained by hot neutral beam injection. The neutrons from the fusion plasma drive the fission assembly which is composed of natural uranium carbide fuel rods clad with stainless steel and is cooled by helium. It was shown how the reactor can be built using essentially present day construction technology and how the uranium bearing blanket modules can be routinely changed to allow separation of the bred fissile fuel of which approximately 1200 kg of plutonium are produced each year along with the approximately 750 MW of electricity. (U.S.)

Neutronics analysis of water-cooled energy production blanket for a fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Jiang Jieqiong; Wang Minghuang; Chen Zhong; Qiu Yuefeng; Liu Jinchao; Bai Yunqing; Chen Hongli; Hu Yanglin

2010-01-01

Neutronics calculations were performed to analyse the parameters of blanket energy multiplication factor (M) and tritium breeding ratio (TBR) in a fusion-fission hybrid reactor for energy production named FDS (Fusion-Driven hybrid System)-EM (Energy Multiplier) blanket. The most significant and main goal of the FDS-EM blanket is to achieve the energy gain of about 1 GWe with self-sustaining tritium, i.e. the M factor is expected to be ∼90. Four different fission materials were taken into account to evaluate M in subcritical blanket: (i) depleted uranium, (ii) natural uranium, (iii) enriched uranium, and (iv) Nuclear Waste (transuranic from 33 000 MWD/MTU PWR (Pressurized Water Reactor) and depleted uranium) oxide. These calculations and analyses were performed using nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library) and a home-developed code VisualBUS. The results showed that the performance of the blanket loaded with Nuclear Waste was most attractive and it could be promising to effectively obtain tritium self-sufficiency and a high-energy multiplication.

Conceptual design of power conversion system for a fusion power reactor with self-cooled LiPb-blanket. EFDA Task TW2-TRP-PPCS12 - Deliverable 4

International Nuclear Information System (INIS)

Vieider, Gottfried

2002-05-01

For FPRs with self-cooled LiPb-blanket and He-cooled first wall and divertor a conceptual design of the power conversion system is developed with emphasis on component feasibility, safety, reliability and thermal efficiency. The resulting power conversion system with a steam turbine is based on proven technology for Na- and He-cooled fission reactors and is assessed to yield an overall net thermal plant efficiency of ∼40 % provided the high primary coolant temperatures of ∼700 deg C can be achieved. The required complexity of the five linked cooling systems can be expected to influence plant cost and reliability

Method of operating water cooled reactor with blanket

International Nuclear Information System (INIS)

Suzuki, Katsuo.

1988-01-01

Purpose: To increase the production amount of fissionable plutonium by increasing the burnup degree of blanket fuels in a water cooled reactor with blanket. Method: Incore insertion assemblies comprising water elimination rods, fertile material rods or burnable poison rods are inserted to those fuel assemblies at the central portion of the reactor core that are situated at the positions not inserted with control rods in the earlier half of the operation cycle, while the incore reactor insertion assemblies are withdrawn at the latter half of the operation cycle of a nuclear reactor. As a result, it is possible to increase the power share of the blanket fuels and increase the fuel burnup degree to thereby increase the production amount of fissionable plutonium. Furthermore, at the initial stage of the cycle, the excess reactivity of the reactor can be suppressed to decrease the reactivity control share on the control rod. At the final stage of the cycle, the excess reactivity of the reactor core can be increased to improve the cycle life. (Kamimura, M.)

Cooling System Design Options for a Fusion Reactor

Science.gov (United States)

Natalizio, Antonio; Collén, Jan; Vieider, Gottfried

1997-06-01

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

Design of a Mechanical NaK Pump for Fission Space Power Systems

Science.gov (United States)

Mireles, Omar R.; Bradley, David; Godfroy, Thomas

2010-01-01

Alkali liquid metal cooled fission reactor concepts are under development for mid-range spaceflight power requirements. One such concept utilizes a sodium-potassium eutectic (NaK) as the primary loop working fluid. Traditionally, linear induction pumps have been used to provide the required flow and head conditions for liquid metal systems but can be limited in performance. This paper details the design, build, and check-out test of a mechanical NaK pump. The pump was designed to meet reactor cooling requirements using commercially available components modified for high temperature NaK service.

Passive cooling systems in power reactors

International Nuclear Information System (INIS)

Aharon, J.; Harrari, R.; Weiss, Y.; Barnea, Y.; Katz, M.; Szanto, M.

1996-01-01

This paper reviews several R and D activities associated with the subject of passive cooling systems, conducted by the N.R.C.Negev thermohydraulic group. A short introduction considering different types of thermosyphons and their applications is followed by a detailed description of the experimental work, its results and conclusions. An ongoing research project is focused on the evaluation of the external dry air passive containment cooling system (PCCS) in the AP-600 (Westinghouse advanced pressurized water reactor). In this context some preliminary theoretical results and planned experimental research are for the fature described

Cooling system upon reactor isolation

International Nuclear Information System (INIS)

Yamamoto, Kohei; Oda, Shingo; Miura, Satoshi

1992-01-01

A water level indicator for detecting the upper limit value for a range of using a suppression pool and a thermometer for detecting the temperature of water at the cooling water inlet of an auxiliary device are disposed. When a detection signal is intaken and the water level in the suppression pool reach the upper limit value for the range of use, a secondary flow rate control value is opened and a primary flow rate control valve is closed. When the temperature of the water at the cooling water inlet of the auxiliary device reaches the upper limit value, the primary and the secondary flow rate control valves are opened. During a stand-by state, the first flow rate control valve is set open and the secondary flow rate control valve is set closed respectively. After reactor isolation, if a reactor water low level signal is received, an RCIC pump is actuated and cooling water is sent automatically under pressure from a condensate storage tank to the reactor and the auxiliary device requiring coolants by way of the primary flow rate control valve. Rated flow rate is ensured in the reactor and cooling water of an appropriate temperature can be supplied to the auxiliary device. (N.H.)

Progress on the conceptual design of a mirror hybrid fusion--fission reactor

International Nuclear Information System (INIS)

Moir, R.W.; Lee, J.D.; Burleigh, R.J.

1975-01-01

A conceptual design study was made of a fusion-fission reactor for the purpose of producing fissile material and electricity. The fusion component is a D-T plasma confined by a pair of magnetic mirror coils in a Yin-Yang configuration and is sustained by neutral beam injection. The neutrons from the fusion plasma drive the fission assembly which is composed of natural uranium carbide fuel rods clad with stainless steel and helium cooled. It was shown conceptually how the reactor might be built using essentially present-day technology and how the uranium-bearing blanket modules can be routinely changed to allow separation of the bred fissile fuel

World Energy Data System (WENDS). Volume XI. Nuclear fission program summaries

International Nuclear Information System (INIS)

1979-06-01

Brief management and technical summaries of nuclear fission power programs are presented for nineteen countries. The programs include the following: fuel supply, resource recovery, enrichment, fuel fabrication, light water reactors, heavy water reactors, gas cooled reactors, breeder reactors, research and test reactors, spent fuel processing, waste management, and safety and environment

AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors

International Nuclear Information System (INIS)

1990-01-01

1 - Description of program or function: AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations. 2 - Method of solution: AUS98 is a modular system in which the modules are complete programs linked by a path given in the input stream. A simple path is simply a sequence of modules, but the path is actually pre-processed and compiled using the Fortran 77 compiler. This provides for complex module linking if required. Some of the modules included in AUS98 are: MIRANDA Cross-section generation in a multi-region resonance subgroup calculation and preliminary group condensation. ANAUSN One-dimensional discrete ordinates calculation. ICPP Isotropic collision probability calculation in one dimension and for rod clusters. POW3D Multi-dimensional neutron diffusion calculation including feedback-free kinetics. AUSIDD One-dimensional diffusion calculation. EDITAR Reaction-rate editing and group collapsing following a transport calculation. CHAR Lattice and global burnup calculation. MICBURN Control of global burnup

Core test reactor shield cooling system analysis

International Nuclear Information System (INIS)

Larson, E.M.; Elliott, R.D.

1971-01-01

System requirements for cooling the shield within the vacuum vessel for the core test reactor are analyzed. The total heat to be removed by the coolant system is less than 22,700 Btu/hr, with an additional 4600 Btu/hr to be removed by the 2-inch thick steel plate below the shield. The maximum temperature of the concrete in the shield can be kept below 200 0 F if the shield plug walls are kept below 160 0 F. The walls of the two ''donut'' shaped shield segments, which are cooled by the water from the shield and vessel cooling system, should operate below 95 0 F. The walls of the center plug, which are cooled with nitrogen, should operate below 100 0 F. (U.S.)

Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

Science.gov (United States)

Tsibulskiy, V. F.; Andrianova, E. A.; Davidenko, V. D.; Rodionova, E. V.; Tsibulskiy, S. V.

2017-12-01

A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium-plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

World Energy Data System (WENDS). Volume XI. Nuclear fission program summaries

Energy Technology Data Exchange (ETDEWEB)

1979-06-01

Brief management and technical summaries of nuclear fission power programs are presented for nineteen countries. The programs include the following: fuel supply, resource recovery, enrichment, fuel fabrication, light water reactors, heavy water reactors, gas cooled reactors, breeder reactors, research and test reactors, spent fuel processing, waste management, and safety and environment. (JWR)

System for cooling the containment vessel of a nuclear reactor

International Nuclear Information System (INIS)

Costes, Didier.

1982-01-01

The invention concerns a post-accidental cooling system for a nuclear reactor containment vessel. This system includes in series a turbine fed by the moist air contained in the vessel, a condenser in which the air is dried and cooled, a compressor actuated by the turbine and a cooling exchanger. The cold water flowing through the condenser and in the exchanger is taken from a tank outside the vessel and injected by a pump actuated by the turbine. The application is for nuclear reactors under pressure [fr

Gas-cooled nuclear reactor

International Nuclear Information System (INIS)

1974-01-01

The invention aims at simplying gas-cooled nuclear reactors. For the cooling gas, the reactor is provided with a main circulation system comprising one or several energy conversion main groups such as gas turbines, and an auxiliary circulation system comprising at least one steam-generating boiler heated by the gas after its passage through the reactor core and adapted to feed a steam turbine with motive steam. The invention can be applied to reactors the main groups of which are direct-cycle gas turbines [fr

Radiation shielding for fission reactors

Energy Technology Data Exchange (ETDEWEB)

Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Laboratory, Tokyo (Japan)

2000-03-01

Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium. (author)

Power deposition distribution in liquid lead cooled fission reactors and effects on the reactor thermal behaviour

International Nuclear Information System (INIS)

Cevolani, S.; Nava, E.; Burn, K. W.

2001-01-01

In the framework of an ADS study (Accelerator Driven System, a reactor cooled by a lead bismuth alloy) the distribution of the deposited energy between the fuel, coolant and structural materials was evaluated by means of Monte Carlo calculations. The energy deposition in the coolant turned out to be about four percent of the total deposited energy. In order to study this effect, further calculations were performed on water and sodium cooled reactors. Such an analysis showed, for both coolant materials, a much lower heat deposition, about one percent. Based on such results, a thermohydraulic analysis was performed in order to verify the effect of this phenomenon on the fuel assembly temperature distribution. The main effect of a significant fraction of energy deposition in the coolant is concerned with the decrease of the fuel pellet temperature. As a consequence, taking into account this effect allows to increase the possibilities of optimization at the disposal of the designer [it

Preconceptual design and analysis of a solid-breeder blanket test in an existing fission reactor

International Nuclear Information System (INIS)

Deis, G.A.; Hsu, P.Y.; Watts, K.D.

1983-01-01

Preconceptual design and analysis have been performed to examine the capabilities of a proposed fission-based test of a water-cooled Li 2 O blanket concept. The mechanical configuration of the test piece is designed to simulate a unit cell of a breeder-outside-tube concept. This test piece will be placed in a fission test reactor, which provides an environment similar to that in a fusion reactor. The neutron/gamma flux from the reactor produces prototypical power density, tritium production rates, and operating temperatures and stresses. Steady-state tritium recovery from the test piece can be attained in short-duration (5-to-6-day) tests. The capabilities of this test indicate that fission-based testing can provide important near-term engineering information to support the development of fusion technology

Thermochemical Analysis of Gas-Cooled Reactor Fuels Containing Am and Pu Oxides

International Nuclear Information System (INIS)

Lindemer, T.B.

2002-01-01

Literature values and estimated data for the thermodynamics of the actinide oxides and fission products are applied to explain the chemical behavior in gas-cooled-reactor fuels. Emphasis is placed on the Am-O-C and Pu-O-C systems and the data are used to plot the oxygen chemical potential versus temperature of solid-solid and solid-gas equilibria. These results help explain observations of vaporization in Am oxides, nitrides, and carbides and provide guidance for the ceramic processing of the fuels. The thermodynamic analysis is then extended to the fission product systems and the Si-C-O system. Existing data on oxygen release (primarily as CO) as a function of burnup in the thoria-urania fuel system is reviewed and compared to values calculated from thermodynamic data. The calculations of oxygen release are then extended to the plutonia and americia fuels. Use of ZrC not only as a particle coating that may be more resistant to corrosion by Pd and other noble-metal fission products, but also as a means to getter oxygen released by fission is discussed

Preliminary design of a borax internal core-catcher for a gas cooled fast reactor

International Nuclear Information System (INIS)

Dalle Donne, M.; Dorner, S.; Schumacher, G.

1976-09-01

Preliminary thermal calculations show that a core-catcher appears to be feasible, which is able to cope with the complete meltdown of the core and blankets of a 1,000 MWe GCFR. This core-catcher is based on borax (Na 2 B 4 O 7 ) as dissolving material of the oxide fuel and of the fission products occuring in oxide form. The borax is contained in steel boxes forming a 2.1 meter thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel, just underneath the reactor core. The fission products are dispersed in the pool formed by the liquid borax. The heat power density in the pool is conveniently reduced and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system. (orig.) [de

Fast reactor cooled by supercritical light water

Energy Technology Data Exchange (ETDEWEB)

Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

2001-09-01

This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

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

International Nuclear Information System (INIS)

Monteith, H.C.

1978-10-01

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

Space Fission System Test Effectiveness

International Nuclear Information System (INIS)

Houts, Mike; Schmidt, Glen L.; Van Dyke, Melissa; Godfroy, Tom; Martin, James; Bragg-Sitton, Shannon; Dickens, Ricky; Salvail, Pat; Harper, Roger

2004-01-01

Space fission technology has the potential to enable rapid access to any point in the solar system. If fission propulsion systems are to be developed to their full potential, however, near-term customers need to be identified and initial fission systems successfully developed, launched, and utilized. One key to successful utilization is to develop reactor designs that are highly testable. Testable reactor designs have a much higher probability of being successfully converted from paper concepts to working space hardware than do designs which are difficult or impossible to realistically test. ''Test Effectiveness'' is one measure of the ability to realistically test a space reactor system. The objective of this paper is to discuss test effectiveness as applied to the design, development, flight qualification, and acceptance testing of space fission systems. The ability to perform highly effective testing would be particularly important to the success of any near-term mission, such as NASA's Jupiter Icy Moons Orbiter, the first mission under study within NASA's Project Prometheus, the Nuclear Systems Program

Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

Science.gov (United States)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Roberts, Jeremy A.; Unruh, Troy C.; McGregor, Douglas S.

2018-01-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of

Thermal-hydraulic simulation and analysis of Research Reactor Cooling Systems

International Nuclear Information System (INIS)

EL Khatib, H.H.A.

2013-01-01

The objective of the present study is to formulate a model to simulate the thermal hydraulic behavior of integrated cooling system in a typical material testing reactor (MTR) under loss of ultimate heat sink, the model involves three interactively coupled sub-models for reactor core, heat exchanger and cooling tower. The developed model predicts the temperature profiles in addition it predicts inlet and outlet temperatures of the hot and cold stream as well as the heat exchangers and cooling tower. The model is validated against PARET code for steady-state operation and also verified by the reactor operational records, and then the model is used to simulate the thermal-hydraulic behavior of the reactor under a loss of ultimate heat sink. The simulation is performed for two operational regimes named regime I of (11 MW) thermal power and three operated cooling tower cells and regime II of (22 MW) thermal power and six operated cooling tower cells. In regime I, the simulation is performed for 1, 2 and 3 cooling tower failed cells while in regime II, it is performed for 1, 2, 3, 4, 5 and 6 cooling tower failed cells. The safety action is conducted by the reactor protection system (RPS) named power reduction safety action, it is triggered to decrease the reactor power by amount of 20% of the present power when the water inlet temperature to the core reaches 43 degree C and a scram (emergency shutdown) is triggered in case of the inlet temperature reaches 44 degree C. The model results are analyzed and discussed. The temperature profiles of fuel, clad and coolant are predicted during transient where its maximum values are far from thermal hydraulic limits.

In-service inspections of the reactor cooling system of pressurized water reactors

International Nuclear Information System (INIS)

Fuerste, W.; Hohnerlein, G.; Werden, B.

1982-01-01

In order to guarantee constant safety of the components of the reactor cooling system, regular in-service inspections are carried out after commissioning of the nuclear power plant. This contribution is concerned with the components of the reactor cooling system, referring to the legal requirements, safety-related purposes and scope of the in-service inspections during the entire period of operation of a nuclear power plant. Reports are made with respect to type, examination intervals, examination technique, results and future development. The functional tests which are carried out within the scope of the in-service inspections are not part of this contribution. (orig.) [de

Gas-cooled reactor power systems for space

International Nuclear Information System (INIS)

Walter, C.E.

1987-01-01

Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system

HAC and fission reactors

International Nuclear Information System (INIS)

Fujiwara, I.; Moriyama, H.; Tachikawa, E.

1984-01-01

In the fission process, newly formed fission products undergo hot atom reactions due to their energetic recoil and abnormal positive charge. The hot atom reactions of the fission products are usually accompanied by secondary effects such as radiation damage, especially in condensed phase. For reactor safety it is valuable to know the chemical behaviour and the release behaviour of these radioactive fission products. Here, the authors study the chemical behaviour and the release behaviour of the fission products from the viewpoint of hot atom chemistry (HAC). They analyze the experimental results concerning fission product behaviour with the help of the theories in HAC and other neighboring fields such as radiation chemistry. (Auth.)

Preliminary neutronics calculation of fusion-fission hybrid reactor breeding spent fuel assembly

International Nuclear Information System (INIS)

Ma Xubo; Chen Yixue; Gao Bin

2013-01-01

The possibility of using the fusion-fission hybrid reactor breeding spent fuel in PWR was preliminarily studied in this paper. According to the fusion-fission hybrid reactor breeding spent fuel characteristics, PWR assembly including fusion-fission hybrid reactor breeding spent fuel was designed. The parameters such as fuel temperature coefficient, moderator temperature coefficient and their variation were investigated. Results show that the neutron properties of uranium-based assembly and hybrid reactor breeding spent fuel assembly are similar. The design of this paper has a smaller uniformity coefficient of power at the same fissile isotope mass percentage. The results will provide technical support for the future fusion-fission hybrid reactor and PWR combined with cycle system. (authors)

Study of advanced fission power reactor development for the United States. Volume II

International Nuclear Information System (INIS)

1976-01-01

This report presents the results of a multi-phase research study which had as its objective the comparative study of various advanced fission reactors and evaluation of alternate strategies for their development in the USA through the year 2020. By direction from NSF, ''advanced'' reactors were defined as those which met the dual requirements of (1) offering a significant improvement in fissile fuel utilization as compared to light-water reactors and (2) currently receiving U.S. Government funding. (A detailed study of the LMFBR was specifically excluded, but cursory baseline data were obtained from ERDA sources.) Included initially were the High-Temperature Gas-Cooled Reactor (HTGR), Gas-Cooled Fast Reactor (GCFR), Molten Salt Reactor (MSR), and Light-Water Breeder Reactor (LWBR). Subsequently, the CANDU Heavy Water Reactor (HWR) was included for comparison due to increased interest in its potential. This volume presents the reasoning process and analytical methods utilized to arrive at the conclusions for the overall study

Control device for can failures of liquid cooled nuclear reactor fuel elements

International Nuclear Information System (INIS)

Althaus, D.; Mohm, F.; Nyhof, M.

1974-01-01

Checking of the fuel or breeding elements of, e.g., sodium-cooled reactors is done by detecting fission products in the coolant, with a flushing gas like argon removing the fission products from the coolant and carrying them to a detector system. In order to increase the escape rate of the fission products from the elements, these are lifted by a hoisting unit into a pit reaching down below the coolant level and are then heated. Heating is achieved by the decay heat or by an additional heating in the receiving pit which at this point is thermically insulated from the exterior. The flushing gas is blown radially into the receiving pit from below. The rising bubbles take along the fission products to a scintillation counter mounted on the head of the receiving vessel. This vessel may have double walls with coolant flowing through the interspace. (DG) [de

Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Matsuura, Hideaki, E-mail: mat@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Nakaya, Hiroyuki; Nakao, Yasuyuki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki 311-1393 (Japan); Nishikawa, Masabumi [Graduate School of Engineering Science, Kyushu University, 6-10-1 Hakozaki, Fukuoka 812-8581 (Japan)

2013-10-15

Highlights: • The performance of a gas-cooled reactor as a tritium production system was studied. • A continuous tritium recovery using helium gas was considered. • Gas-cooled reactors with 3 GW output in all can produce ∼6 kg of tritium in a year • Performance of the system was examined for Li{sub 4}SiO{sub 4}, Li{sub 2}TiO{sub 3} and LiAlO{sub 2} compounds. -- Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year.

Indian programme on molten salt cooled nuclear reactors

International Nuclear Information System (INIS)

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

2013-01-01

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

Steam generators in indirect-cycle water-cooled reactors

International Nuclear Information System (INIS)

Fajeau, M.

1976-01-01

In the indirect cycle water-cooled nuclear reactors, the steam generators are placed between the primary circuit and the turbine. They act both as an energy transmitter and as a leaktigh barrier against fission or corrosion products. Their study is thus very important from a performance and reliability point of view. Two main types are presented here: the U-tube and the once-through steam generators [fr

Gas-cooled reactor for space power systems

International Nuclear Information System (INIS)

Walter, C.E.; Pearson, J.S.

1987-05-01

Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors

Radioactivities evaluation code system for high temperature gas cooled reactors during normal operation

International Nuclear Information System (INIS)

Ogura, Kenji; Morimoto, Toshio; Suzuki, Katsuo.

1979-01-01

A radioactivity evaluation code system for high temperature gas-cooled reactors during normal operation was developed to study the behavior of fission products (FP) in the plants. The system consists of a code for the calculation of diffusion of FPs in fuel (FIPERX), a code for the deposition of FPs in primary cooling system (PLATO), a code for the transfer and emission of FPs in nuclear power plants (FIPPI-2), and a code for the exposure dose due to emitted FPs (FEDOSE). The FIPERX code can calculate the changes in the course of time FP of the distribution of FP concentration, the distribution of FP flow, the distribution of FP partial pressure, and the emission rate of FP into coolant. The amount of deposition of FPs and their distribution in primary cooling system can be evaluated by the PLATO code. The FIPPI-2 code can be used for the estimation of the amount of FPs in nuclear power plants and the amount of emitted FPs from the plants. The exposure dose of residents around nuclear power plants in case of the operation of the plants is calculated by the FEDOSE code. This code evaluates the dose due to the external exposure in the normal operation and in the accident, and the internal dose by the inhalation of radioactive plume and foods. Further studies of this code system by the comparison with the experimental data are considered. (Kato, T.)

Failure Mode and Effects Analysis (FMEA) of the Emergency Core Cooling System (ECCS) for a Westinghouse type 312, three loop pressurized water reactor

International Nuclear Information System (INIS)

Shopsky, W.E.

1977-01-01

The Emergency Core Cooling System (ECCS) is a Safeguards System designed to cool the core in the unlikely event of a Loss-of-Coolant Accident (LOCA) in the primary reactor coolant system as well as to provide additional shutdown capability following a steam break accident. The system is designed for a high reliability of providing emergency coolant and shutdown reactivity to the core for all anticipated occurrences of such accidents. The ECCS by performing its intended function assures that fuel and clad damage is minimized during accident conditions thus reducing release of fission products from the fuel. The ECCS is designed to perform its function despite sustaining a single failure by the judicious use of equipment and flow path redundancy within and outside the containment structure. By the use of an analytic tool, a Failure Mode and Effects Analysis (FMEA), it is shown that the ECCS is in compliance with the Single Failure Criterion established for active failures of fluid systems during short and long term cooling of the reactor core following a LOCA or steam break accident. An analysis was also performed with regards to passive failure of ECCS components during long-term cooling of the core following an accident. The design of the ECCS was verified as being able to tolerate a single passive failure during long-term cooling of the reactor core following an accident. The FMEA conducted qualitatively demonstrates the reliability of the ECCS (concerning active components) to perform its intended safety function

A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

International Nuclear Information System (INIS)

Koch, M.; Kazimi, M.S.

1991-04-01

Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed

A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

Energy Technology Data Exchange (ETDEWEB)

Koch, M.; Kazimi, M.S.

1991-04-01

Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

Natural circulating passive cooling system for nuclear reactor containment structure

Science.gov (United States)

Gou, Perng-Fei; Wade, Gentry E.

1990-01-01

A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

System design study of small lead-bismuth cooled reactor

International Nuclear Information System (INIS)

Chikazawa, Yoshitaka; Hori, Toru; Konomura, Mamoru

2003-07-01

In phase II of the feasibility study of JNC, we will make a concept of a dispersion power source reactor with various requirements, such as economical competitiveness and safety. In the study of a small lead-bismuth cooled reactor, a concept whose features are long life core, inherent safety, natural convection of cooling system and steam generators in the reactor vessel has been designed since 2000. The investigations which have been done in 2002 are shown as follows; Safety analysis of UTOP considering uncertainty of reactivity. Possibility of reduction of number of control rods. Estimation of construction cost. Transient analyses of UTOP have been done in considering uncertainty of reactivity in order to show the inherent safety in the probabilistic method. And the inherent safety in UTOP is realized under the condition of considering uncertainty. Transient analyses of UTOP with various numbers of control rods have been done and it is suggested that there is possibility of reduction of the number of control rods considering accident managements. The method of cost estimation is a little modified. The cost of reactor vessel is estimated from that of medium sized lead-bismuth cooled reactor and the estimation of a purity control system is by coolant volume flow rate. The construction cost is estimated 850,000yen/kWe. (author)

Fission product chemistry in severe nuclear reactor accidents

International Nuclear Information System (INIS)

Nichols, A.L.

1990-09-01

A specialist's meeting was held at JRC-Ispra from 15 to 17 January 1990 to review the current understanding of fission-product chemistry during severe accidents in light water reactors. Discussions focussed on the important chemical phenomena that could occur across the wide range of conditions of a damaged nuclear plant. Recommendations for future chemistry work were made covering the following areas: (a) fuel degradation and fission-product release, (b) transport and attenuation processes in the reactor coolant system, (c) containment chemistry (iodine behaviour and core-concrete interactions)

observer-based diagnostics and monitoring of vibrations in nuclear reactor core cooling system

International Nuclear Information System (INIS)

Siry, S.A K.

2007-01-01

analysis and diagnostics of vibration in industrial systems play a significant rule to prevent severe severe damages . drive shaft vibration is a complicated phenomenon composed of two independent forms of vibrations, translational and torsional. translational vibration measurements in case of the reactor core cooling system are introduced. the system under study consists of the three phase induction motor, flywheel, centrifugal pump, and two coupling between motor-flywheel, and flywheel-pump. this system structure is considered to be one where the blades are pegged into the discs fitting into the shafts. a non-linear model to simulate vibration in the reactor core cooling system will be introduced. simulation results of an operating reactor core cooling system using the actual parameters will be presented to validate the accuracy and reliability of the proposed analytical method the accuracy in analyzing the results depends on the system model. the shortcomings of the conventional model will be avoided through the use of that accurate nonlinear model which improve the simulation of the reactor core cooling system

Conceptual Analysis of Fission Fragment Magnetic Collimator Reactors

International Nuclear Information System (INIS)

Tsvetkov, Pavel V.; Parish, Theodore A.

2002-01-01

As part of the current research work within the US DOE NERI Direct Electricity Conversion (DEC) Project on methods for utilizing direct electricity conversion in nuclear reactors, a detailed study of a Fission Fragment Magnetic Collimator Reactor (FFMCR) has been performed. The FFMCR concept is an advanced DEC system that combines advantageous design solutions proposed for application in both fission and fusion reactors. The present study was focused on determining the electrical efficiency and other important operational aspects of the FFMCR concept. In principle, acceptable characteristics have been demonstrated, and results obtained are presented in the paper. Technological visibility of the FFMCR concept and required further design development are discussed. Preliminary characteristics of the promising design are outlined. (authors)

Neutronics issues in fusion-fission hybrid reactor design

International Nuclear Information System (INIS)

Liu Chengan

1995-01-01

The coupled neutron and γ-ray transport equations and nuclear number density equations, and its computer program systems concerned in fusion-fission hybrid reactor design are briefly described. The current status and focal point for coming work of nuclear data used in fusion reactor design are explained

Emergency cooling system for a liquid metal cooled reactor

International Nuclear Information System (INIS)

Murata, Ryoichi; Fujiwara, Toshikatsu.

1980-01-01

Purpose: To suitably cool liquid metal as coolant in emergency in a liquid metal cooled reactor by providing a detector for the pressure loss of the liquid metal passing through a cooling device in a loop in which the liquid metal is flowed and communicating the detector with a coolant flow regulator. Constitution: A nuclear reactor is stopped in nuclear reaction by control element or the like in emergency. If decay heat is continuously generated for a while and secondary coolant is insufficiently cooled with water or steam flowed through a steam and water loop, a cooler is started. That is, low temperature air is supplied by a blower through an inlet damper to the cooler to cool the secondary coolant flowed into the cooler through a bypass pipe so as to finally safely stop an entire plant. Since the liquid metal is altered in its physical properties by the temperature at this time, it is detected to regulate the opening of the valve of the damper according to the detected value. (Sekiya, K.)

Transmutation of long-lived fission product (137Cs, 90Sr) by a reactor-accelerator system

International Nuclear Information System (INIS)

Toyama, Shin-ichi; Takashita, Hirofumi; Konashi, Kenji; Sasao, Nobuyuki; Sato, Isamu.

1990-01-01

The report discusses the transmutation of long-lived fission products by a reactor and accelerator. It is important to take some criteria into consideration in transmutation disposal. To satisfy the criteria, a combined system of a reactor and an accelerator is proposed for the transmutation. An outline of the transmutation reactor and the accelerator is presented. The transmutation reactor has the ability to transmute a large quantity of fission products. However, it is desirable to have a high transmutation rate as well as a large disposal ability. Besides the transmutation property, it is necessary to investigate the physics of the transmutation reactor such as nuclear characteristics and burnup properties in order to obtain the most suitable, high performance core concept. A study on those properties is also presented. A high power accelerator is required for the transmutation. So a test linac is developed to accelerate high intensity beams. (N.K.)

The electronuclear cycle: from fission to new reactor systems

International Nuclear Information System (INIS)

Belier, G.; Cugnon, J.; Lapoux, V.; Liatard, E.; Porquet, Marie-Genevieve; Rudolf, G.

2006-09-01

The Joliot Curie School trains each year, and since 1981, PhD students, post-Doctorates and researchers on scientific breakthroughs performed in a topic related to nuclear physics, in a broad range. These proceedings brings together the 11 lectures given at the 2006 session of Joliot Curie School on the topic of the electronuclear cycle: - Fission: from phenomenology to theory (Berger, J.F.); - Physics of nuclear reactors (Baeten, P.); - Data modeling and evaluation (Bauge, E.; Hilaire, S.); - Measurement of cross sections of interest for minor actinides incineration (Jurado, B.); - Spallation data and modelling for hybrid reactors (Boudard, A.); - Nuclear wastes: overview (Billard, I.); - Long living nuclear wastes transmutation processes and feasibility (Varaine, F.); - Hybrid reactors: recent advances for a demonstrator (Billebaud, A.); - Systems of the future and strategy (David, S.); - Non-nuclear energies (Nifenecker, H.); - Fundamental physics with ultracold neutrons (Protasov, K). The last section is a compilation of abstracts of presentations given by Young searchers' (Young searchers' seminars)

An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Bucknor, M.; Farmer, M.; Grabaskas, D.

2017-06-26

The U.S. Nuclear Regulatory Commission has stated that mechanistic source term (MST) calculations are expected to be required as part of the advanced reactor licensing process. A recent study by Argonne National Laboratory has concluded that fission product scrubbing in sodium pools is an important aspect of an MST calculation for a sodium-cooled fast reactor (SFR). To model the phenomena associated with sodium pool scrubbing, a computational tool, developed as part of the Integral Fast Reactor (IFR) program, was utilized in an MST trial calculation. This tool was developed by applying classical theories of aerosol scrubbing to the decontamination of gases produced as a result of postulated fuel pin failures during an SFR accident scenario. The model currently considers aerosol capture by Brownian diffusion, inertial deposition, and gravitational sedimentation. The effects of sodium vapour condensation on aerosol scrubbing are also treated. This paper provides details of the individual scrubbing mechanisms utilized in the IFR code as well as results from a trial mechanistic source term assessment led by Argonne National Laboratory in 2016.

Concept of safe tank-type water cooled and moderated reactor with HTGR microparticle fuel compacts

International Nuclear Information System (INIS)

Gol'tsev, A.O.; Kukharkin, N.E.; Mosevitskij, I.S.; Ponomarev-Stepnoj, N.N.; Popov, S.V.; Udyanskij, Yu.N.; Tsibul'skij, V.F.

1993-01-01

Concept of safe tank-type water-cooled and moderated reactor on the basis of HTGR fuel microparticles which enable to avoid environment contamination with radioactive products under severe accidents, is proposed. Results of neutron-physical and thermal-physical studies of water cooled and moderated reactor with HTGR microparticle compacts are presented. Characteristics of two reactors with thermal power of 500 and 1500 MW are indicated within the concept frames. The reactor behaviour under severe accident connected with complete loss of water coolant is considered. It is shown that under such an accident the fission products release from fuel microparticles does not occur

High-temperature gas-cooled reactor safety studies. Progress report for January 1, 1974--June 30, 1975

International Nuclear Information System (INIS)

Cole, T.E.; Sanders, J.P.; Kasten, P.R.

1977-07-01

Progress is reported in the following areas: systems and safety analysis; fission product technology; primary coolant technology; seismic and vibration technology; confinement components; primary system materials technology; safety instrumentation; loss of flow accident analysis using HEATUP code; use of coupled-conduction-convection model for core thermal analysis; development of multichannel conduction-convection program HEXEREI; cooling system performance after shutdown; core auxiliary cooling system performance; development of FLODIS code; air ingress into primary systems following DBDA; performance of PCRV thermal barrier cover plates; temperature limits for fuel particle coating failure; tritium distribution and release in HTGR; energy release to PCRV during DBDA; and mathematical models for HTGR reactor safety studies

Fission reactor container

International Nuclear Information System (INIS)

Akagawa, Katsuhiko.

1991-01-01

Cooling water is sent without using dynamic equipments upon loss of coolants accident in a pressure vessel by improving an arrangement of a nuclear reactor pressure vessel. That is, a containing space is formed at the center of a suppression chamber for storing cooling water while being partitioned with each other, in which the pressure vessel is placed. Further, a water reservoir is formed above the pressure vessel. Then a water discharge pipe is connected to the reservoir for submerging the stored water over the pressure vessel upon occurrence of loss of coolants accident. Further, a water injection pipe is disposed between the pressure suppression chamber and the pressure vessel for injecting the cooling water in the pressure suppression chamber to the reactor core of the pressure vessel by the difference of a water head upon loss of coolants accident. With such a constitution, the pressure vessel has high earthquake proofness. Further, upon loss of coolants accident of the pressure vessel, the cooling water in the reservoir is discharged to submerge and cool the pressure vessel efficiently. Further, the reactor core of the pressure vessel can certainly be cooled by the cooling water of the pressure suppression chamber without relying on dynamic equipments. (I.S.)

A long term radiological risk model for plutonium-fueled and fission reactor space nuclear system

International Nuclear Information System (INIS)

Bartram, B.W.; Dougherty, D.K.

1987-01-01

This report describes the optimization of the RISK III mathematical model, which provides risk assessment for the use of a plutonium-fueled, fission reactor in space systems. The report discusses possible scenarios leading to radiation releases on the ground; distinctions are made for an intact reactor and a dispersed reactor. Also included are projected dose equivalents for various accident situations. 54 refs., 31 figs., 11 tabs

Utilization of fission reactors for fusion engineering testing

International Nuclear Information System (INIS)

Deis, G.A.; Miller, L.G.

1985-01-01

Fission reactors can be used to conduct some of the fusion nuclear engineering tests identified in the FINESSE study. To further define the advantages and disadvantages of fission testing, the technical and programmatic constraints on this type of testing are discussed here. This paper presents and discusses eight key issues affecting fission utilization. Quantitative comparisons with projected fusion operation are made to determine the technical assets and limitations of fission testing. Capabilities of existing fission reactors are summarized and compared with technical needs. Conclusions are then presented on the areas where fission testing can be most useful

A domain-specific analysis system for examining nuclear reactor simulation data for light-water and sodium-cooled fast reactors

International Nuclear Information System (INIS)

Billings, Jay Jay; Deyton, Jordan H.; Forest Hull, S.; Lingerfelt, Eric J.; Wojtowicz, Anna

2015-01-01

Highlights: • Data analysis for high-performance simulations of reactors will be a problem that we address with a new management system. • We describe new input-output libraries for nuclear reactor simulations. • We describe a new user interface for visualizing and analyzing simulation results. • We show the utility of these systems with a 17 × 17 fuel assembly example simulation. • The availability of the code and avenues for collaboration are presented. - Abstract: Building a new generation of fission reactors in the United States presents many technical and regulatory challenges. One important challenge is the need to share and present results from new high-fidelity, high-performance simulations in an easily usable way. Since modern multiscale, multi-physics simulations can generate petabytes of data, they will require the development of new techniques and methods to reduce the data to familiar quantities of interest (e.g., pin powers, temperatures) with a more reasonable resolution and size. Furthermore, some of the results from these simulations may be new quantities for which visualization and analysis techniques are not immediately available in the community and need to be developed. This paper describes a new system for managing high-performance simulation results in a domain-specific way that naturally exposes quantities of interest for light water and sodium-cooled fast reactors. It describes requirements to build such a system and the technical challenges faced in its development at all levels (simulation, user interface, etc.). An example comparing results from two different simulation suites for a single assembly in a light-water reactor is presented, along with a detailed discussion of the system’s requirements and design

Fission rate measurements in fuel plate type assembly reactor cores

International Nuclear Information System (INIS)

Rogers, J.W.

1988-01-01

The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs

High-Temperature Gas-cooled Reactor steam-cycle/cogeneration lead plant reactor vessel: system design description

International Nuclear Information System (INIS)

1983-01-01

The Reactor Vessel System contains the primary coolant inventory within a gas-tight pressure boundary, and provides the necessary flow paths and overpressure protection for this pressure boundary. The Reactor Vessel System also houses the components of the Reactor System, the Heat Transport System, and the Auxiliary Heat Removal System. The scope of the Reactor Vessel System includes the prestressed concrete reactor vessel (PCRV) structure with its reinforcing steel and prestressing components; liners, penetrations, closures, and cooling water tubes attached to the concrete side of the liner; the thermal barrier (insulation) on the primary coolant side of the liner; instrumentation for structural monitoring; and a pressure relief system. Specifications are presented

Plutonium-burn high temperature gas-cooled reactor for 3E+3S

International Nuclear Information System (INIS)

Okamoto, Koji

2015-01-01

The Nuclear Energy Development in Japan is facing a very difficult conditions after Fukushima-Daiichi NPP Accident. Nuclear Energy has strong advantages on 3E, i.e., Energy security, Economical efficiency and Environment. However, people does not believe the Safety 'S' of Nuclear Energy, now. The disadvantage of 'S' overrides the advantages of '3E'. In Nuclear Energy, 'S' is expanded into 3S, i.e., Safety, Security and Safeguards. Especially, the management of Plutonium inventory in Spent Fuel generated by the NPP operation is very important in the viewpoints of non-proliferation. The high-temperature gas cooled reactor (HTGR) is the solution of these disadvantages of '3S' in Nuclear Energy. The fuel of HTGR is composed by 1 mm spherical fuel particle, i.e., TRISO made by fuel, graphite and silicon-carbide. The silicon-carbide can confine the fission products in any conditions of fuel life cycle, i.e., during operation, accidents and disposal for 1 million years. The confinement of the radioactive materials can be confirmed by the TRISO. The HTGR core has strong negative feedback for temperature. So, the fission automatically stopped at the accidental conditions, such as loss of flow and LOCA. Also, the residual heat can be cooled by the radiation heat transfer to reactor vessel wall. The HTGR system usually has passive vessel wall cooling system. When the passive cooling system had been failed, the heat can be transferred to the land by heat conductions, and fuel does not reach the SiC broken temperature. The fission chain reaction has been stopped automatically by negative feedback, i.e., physics. The residual heat had been cooled automatically by radiation. The radioactive materials had been confined automatically by silicon-carbide. The HTGR is superior for 'S' safety. Plutonium can be burned by the HTGR. In the viewpoints of non-proliferation, the fuel should be made by YSZ-PuO 2 , stabilized buffer

Renewal of reactor cooling system of JMTR. Reactor building site

International Nuclear Information System (INIS)

Onoue, Ryuji; Kawamata, Takanori; Otsuka, Kaoru; Sekine, Katsunori; Koike, Sumio; Gorai, Shigeru; Nishiyama, Yutaka; Fukasaku, Akitomi

2012-03-01

The Japan Materials Testing Reactor (JMTR) is a light water moderated and cooled tank-type reactor, and its thermal power is 50 MW. The JMTR is categorized as high flux testing reactors in the world. The JMTR has been utilized for irradiation experiments of nuclear fuels and materials, as well as for radioisotope productions since the first criticality in March 1968 until August 2006. JAEA is decided to refurbish the JMTR as an important fundamental infrastructure to promote the nuclear research and development. And The JMTR refurbishment work is carried out for 4 years from 2007. Before refurbishment work, from August 2006 to March 2007, all concerned renewal facilities were selected from evaluation on their damage and wear in terms of aging. Facilities which replacement parts are no longer manufactured or not likely to be manufactured continuously in near future, are selected as renewal ones. Replace priority was decided with special attention to safety concerns. A monitoring of aging condition by the regular maintenance activity is an important factor in selection of continuous using after the restart. In this report, renewal of the cooling system within refurbishment facilities in the JMTR is summarized. (author)

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

International Nuclear Information System (INIS)

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

2003-02-01

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

Fission gas behaviour in water reactor fuels

International Nuclear Information System (INIS)

2002-01-01

During irradiation, nuclear fuel changes volume, primarily through swelling. This swelling is caused by the fission products and in particular by the volatile ones such as krypton and xenon, called fission gas. Fission gas behaviour needs to be reliably predicted in order to make better use of nuclear fuel, a factor which can help to achieve the economic competitiveness required by today's markets. These proceedings communicate the results of an international seminar which reviewed recent progress in the field of fission gas behaviour in light water reactor fuel and sought to improve the models used in computer codes predicting fission gas release. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors. (author)

Convective cooling in a pool-type research reactor

Science.gov (United States)

Sipaun, Susan; Usman, Shoaib

2016-01-01

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

Conceptual design of a fusion-fission hybrid reactor for transmutation of high level nuclear waste

International Nuclear Information System (INIS)

Qiu, L.J.; Wu, Y.C.; Yang, Y.W.; Wu, Y.; Luan, G.S.; Xu, Q.; Guo, Z.J.; Xiao, B.J.

1994-01-01

To assess the feasibility of the transmutation of long-lived radioactive waste using fusion-fission hybrid reactors, we are studying all the possible types of blanket, including a comparison of the thermal and fast neutron spectrum blankets. Conceptual designs of a small tokamak hybrid blanket with small inventory of actinides and fission products are presented. The small inventory of wastes makes the system safer. The small hybrid reactor system based on a fusion core with experimental parameters to be realized in the near future can effectively transmute actinides and fission products at a neutron wall loading of 1MWm -2 . An innovative energy system is also presented, including a fusion driver, fuel breeder, high level waste transmuter, fission reactor and so on. An optimal combination of all types of reactor is proposed in the system. ((orig.))

System Design of a Supercritical CO_2 cooled Micro Modular Reactor

International Nuclear Information System (INIS)

Kim, Seong Gu; Cho, Seongkuk; Yu, Hwanyeal; Kim, Yonghee; Jeong, Yong Hoon; Lee, Jeong Ik

2014-01-01

Small modular reactor (SMR) systems that have advantages of little initial capital cost and small restriction on construction site are being developed by many research organizations around the world. Existing SMR concepts have the same objective: to achieve compact size and a long life core. Most of small modular reactors have much smaller size than the large nuclear power plant. However, existing SMR concepts are not fully modularized. This paper suggests a complete modular reactor with an innovative concept for reactor cooling by using a supercritical carbon dioxide. The authors propose the supercritical CO_2 Brayton cycle (S-CO_2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the reactor core and PCU in one vessel. A conceptual design of the proposed small modular reactor was developed, which is named as KAIST Micro Modular Reactor (MMR). The supercritical CO_2 Brayton cycle for the S-CO_2 cooled reactor core was optimized and the size of turbomachinery and heat exchanger were estimated preliminary. The nuclear fuel composed with UN was proposed and the core lifetime was obtained from a burnup versus reactivity calculation. Furthermore, a system layout with fully passive safety systems for both normal operation and emergency operation was proposed. (author)

Systems design of direct-cycle supercritical-water-cooled fast reactors

International Nuclear Information System (INIS)

Oka, Yoshiaki; Koshizuka, Seiichi; Jevremovic, Tatjana; Okano, Yashushi

1995-01-01

The system design of a direct-cycle supercritical-water-cooled fast reactor is presented. The supercritical water does not exhibit a change of phase. the recirculation system, steam separator, and dryer of a boiling water reactor (BWR) are unnecessary. Roughly speaking, the reactor pressure vessel and control rods are similar to those of a pressurized water reactor, the containment and emergency core cooling system are similar to a BWR, and the balance of plant is similar to a supercritical-pressure fossil-fired power plant (FPP). the electric power of the fast converter is 1,508 MW(electric). The number of coolant loops is only two because of the high coolant enthalpy. Containment volume is much reduced. The thermal efficiency is improved 24% over a BWR. The coolant void reactivity is negative by placing thin zirconium-hydride layers between seeds and blankets. The power costs would be much reduced compared with those of a light water reactor (LWR) and a liquid-metal fast breeder reactor. The concept is based on the huge amount of experience with the water coolant technology of LWRs and FPPs. The oxidation of stainless steel cladding is avoided by adopting a much lower coolant temperature than that of the FPP

Thermal analysis of CANDU6 moderator system for loss of cooling

International Nuclear Information System (INIS)

Xu Zhen

2012-01-01

The coolant system and moderator system of CANDU6 are independent. The prompt neutrons are moderated as thermal neutrons by the moderator and the continuous nuclear fission in the reactor is maintained. At the same time the moderator system supplies the heat sink for the heat produced by the neutrons moderation. During the in-service maintenance of plant, the standby RCW which will only cool down reactor coolant system operates instead of RCW and can not supply heat sink for moderator system heat exchanger. As the result, the moderator system will lose heat sink during the operation of standby RCW. To estimate the moderator temperature, the thermal analysis of moderator system for loss of cooling was compared with the experiment data and the system failure caused by the temperature raising was evaluated in this paper. (author)

Meltdown reactor core cooling facility

International Nuclear Information System (INIS)

Matsuoka, Tsuyoshi.

1992-01-01

The meltdown reactor core cooling facility comprises a meltdown reactor core cooling tank, a cooling water storage tank situates at a position higher than the meltdown reactor core cooling tank, an upper pipeline connecting the upper portions of the both of the tanks and a lower pipeline connecting the lower portions of them. Upon occurrence of reactor core meltdown, a high temperature meltdown reactor core is dropped on the cooling tank to partially melt the tank and form a hole, from which cooling water is flown out. Since the water source of the cooling water is the cooling water storage tank, a great amount of cooling water is further dropped and supplied and the reactor core is submerged and cooled by natural convection for a long period of time. Further, when the lump of the meltdown reactor core is small and the perforated hole of the meltdown reactor cooling tank is small, cooling water is boiled by the high temperature lump intruding into the meltdown reactor core cooling tank and blown out from the upper pipeline to the cooling water storage tank to supply cooling water from the lower pipeline to the meltdown reactor core cooling tank. Since it is constituted only with simple static facilities, the facility can be simplified to attain improvement of reliability. (N.H.)

Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

International Nuclear Information System (INIS)

Anon.

1981-01-01

The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

Photofission observations in reactor environments using selected fission-product yields

International Nuclear Information System (INIS)

Gold, R.; Ruddy, F.H.; Roberts, J.H.

1982-01-01

A new method for the observation of photofission in reactor environments is advanced. It is based on the in-situ observation of fission product yield. In fact, at a given in-situ reactor location, the fission product yield is simply a weighted linear combination of the photofission product yield, Y/sub gamma/, and the neutron induced fission product yield, Y/sub n. The weight factors arising in this linear combination are the photofission fraction and neutron induced fission fraction, respectively. This method can be readily implemented with established techniques for measuring in-situ reactor fission product yield. For example, one can use the method based on simultaneous irradiation of radiometric (RM) and solid state track recorder (SSTR) fission monitors. The sensitivity and accuracy and current knowledge of fission product yields. Unique advantages of this method for reactor applications are emphasized

Refurbishment of the Primary Cooling System of the Puspati Triga Reactor

Energy Technology Data Exchange (ETDEWEB)

Ramli, S.; Zakaria, M. F.; Masood, Z. [Malaysian Nuclear Agency, Kajang (Malaysia)

2014-08-15

The refurbishment of the 27 year old primary cooling system of the 1 MW PUSPATI TRIGA reactor was completed in April 2010 over an eight month outage. The project was implemented with the dual objective of meeting current user needs as well as a future reactor core power upgrade. Hence the cooling system was partly modernized to cater for a 3 MW{sub th} reactor by installing higher capacity heat exchangers and pumps while maintaining the piping and valve sizes. The old 1 MW tube and shell heat exchanger, which had lost 25% of its heat exchange capacity, was replaced with two 1.5 MW plate type heat exchangers. Several manually operated valves were replaced with motorized units to allow remote operation from the control room. The installed cooling system was flushed with distilled water and then subjected to hydrostatic pressure tests. In the cold run test, the system was operated for an hour for every pump and heat exchanger combination while all operating parameters were checked. In the hot run test, the same was done at four levels of increasing reactor power, and dose measurements were also recorded. The paper gives the design, installation, testing and commissioning details of the project. (author)

Developments and Tendencies in Fission Reactor Concepts

Science.gov (United States)

Adamov, E. O.; Fuji-Ie, Y.

This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

International Nuclear Information System (INIS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-01-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

Emergency cooling system for the PHENIX reactor

International Nuclear Information System (INIS)

Megy, J.M.; Giudicelli, A.G.; Robert, E.A.; Crette, J.P.

Among various engineered safeguards of the reactor plant, the authors describe the protective system designed to remove the decay heat in emergency, in case of complete loss of all normal decay heat removal systems. First the normal decay heat rejection systems are presented. Incidents leading to the loss of these normal means are then analyzed. The protective system and its constructive characteristics designed for emergency cooling and based on two independent and highly reliable circuits entirely installed outside the primary containment vessel are described

CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

Directory of Open Access Journals (Sweden)

Pengcheng Zhao

2016-01-01

Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

Updated comparison of economics of fusion reactors with advanced fission reactors

International Nuclear Information System (INIS)

Delene, J.G.

1990-01-01

The projected cost of electricity (COE) for fusion is compared with that from current and advanced nuclear fission and coal-fired plants. Fusion cost models were adjusted for consistency with advanced fission plants and the calculational methodology and cost factors follow guidelines recommended for cost comparisons of advanced fission reactors. The results show COEs of about 59--74 mills/kWh for the fusion designs considered. In comparison, COEs for future fission reactors are estimated to be in the 43--54 mills/kWh range with coal-fired plant COEs of about 53--69 mills/kWh ($2--3/GJ coal). The principal cost driver for the fusion plants relative to fission plants is the fusion island cost. Although the estimated COEs for fusion are greater than those for fission or coal, the costs are not so high as to preclude fusion's competitiveness as a safe and environmentally sound alternative

Development of failed fuel detection and location system in sodium-cooled large reactor. Sampling method of failed fuels under the slit

International Nuclear Information System (INIS)

Aizawa, Kousuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto

2010-01-01

A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress as an issue of the 'Fast Reactor Cycle Technology Development (FaCT)' project in Japan. JSFR adopts a Selector-Valve mechanism for the failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. In this study, the sampling method for indentifying under-slit failed fuel subassemblies has been demonstrated by water experiments. (author)

Balancing passive and active systems for evolutionary water cooled reactors

International Nuclear Information System (INIS)

Fil, N.S.; Allen, P.J.; Kirmse, R.E.; Kurihara, M.; Oh, S.J.; Sinha, R.K.

1999-01-01

Advanced concepts of the water-cooled reactors are intended to improve safety, economics and public perception of nuclear power. The potential inclusion of new passive means in addition or instead of traditional active systems is being considered by nuclear plant designers to reach these goals. With respect to plant safety, application of the passive means is mainly intended to simplify the safety systems and to improve their reliability, to mitigate the effect of human errors and equipment malfunction. However, some clear drawbacks and the limited experience and testing of passive systems may raise additional questions that have to be addressed in the design process for each advanced reactor. Therefore the plant designer should find a reasonable balance of active and passive means to effectively use their advantages and compensate their drawbacks. Some considerations that have to be taken into account when balancing active/passive means in advanced water-cooled reactors are discussed in this paper. (author)

Assessment of the high temperature fission chamber technology for the French fast reactor program

Energy Technology Data Exchange (ETDEWEB)

Jammes, C.; Filliatre, P.; Geslot, B.; Domenech, T.; Normand, S. [Commissariat a l' Energie Atomique, CEA (France)

2011-07-01

High temperature fission chambers are key instruments for the control and protection of the sodium-cooled fast reactor. First, the developments of those neutron detectors, which are carried out either in France or abroad are reviewed. Second, the French realizations are assessed with the use of the technology readiness levels in order to identify tracks of improvement. (authors)

78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

Science.gov (United States)

2013-10-24

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

New Burnup Calculation System for Fusion-Fission Hybrid System

International Nuclear Information System (INIS)

Isao Murata; Shoichi Shido; Masayuki Matsunaka; Keitaro Kondo; Hiroyuki Miyamaru

2006-01-01

Investigation of nuclear waste incineration has positively been carried out worldwide from the standpoint of environmental issues. Some candidates such as ADS, FBR are under discussion for possible incineration technology. Fusion reactor is one of such technologies, because it supplies a neutron-rich and volumetric irradiation field, and in addition the energy is higher than nuclear reactor. However, it is still hard to realize fusion reactor right now, as well known. An idea of combination of fusion and fission concepts, so-called fusion-fission hybrid system, was thus proposed for the nuclear waste incineration. Even for a relatively lower plasma condition, neutrons can be well multiplied by fission in the nuclear fuel, tritium is thus bred so as to attain its self-sufficiency, enough energy multiplication is then expected and moreover nuclear waste incineration is possible. In the present study, to realize it as soon as possible with the presently proven technology, i.e., using ITER model with the achieved plasma condition of JT60 in JAEA, Japan, a new calculation system for fusion-fission hybrid reactor including transport by MCNP and burnup by ORIGEN has been developed for the precise prediction of the neutronics performance. The author's group already has such a calculation system developed by them. But it had a problem that the cross section libraries in ORIGEN did not have a cross section library, which is suitable specifically for fusion-fission hybrid reactors. So far, those for FBR were approximately used instead in the analysis. In the present study, exact derivation of the collapsed cross section for ORIGEN has been investigated, which means it is directly evaluated from calculated track length by MCNP and point-wise nuclear data in the evaluated nuclear data file like JENDL-3.3. The system realizes several-cycle calculation one time, each of which consists of MCNP criticality calculation, MCNP fixed source calculation with a 3-dimensional precise

Gas cooled fast reactor benchmarks for JNC and Cea neutronic tools assessment

International Nuclear Information System (INIS)

Rimpault, G.; Sugino, K.; Hayashi, H.

2005-01-01

In order to verify the adequacy of JNC and Cea computational tools for the definition of GCFR (gas cooled fast reactor) core characteristics, GCFR neutronic benchmarks have been performed. The benchmarks have been carried out on two different cores: 1) a conventional Gas-Cooled fast Reactor (EGCR) core with pin-type fuel, and 2) an innovative He-cooled Coated-Particle Fuel (CPF) core. Core characteristics being studied include: -) Criticality (Effective multiplication factor or K-effective), -) Instantaneous breeding gain (BG), -) Core Doppler effect, and -) Coolant depressurization reactivity. K-effective and coolant depressurization reactivity at EOEC (End Of Equilibrium Cycle) state were calculated since these values are the most critical characteristics in the core design. In order to check the influence due to the difference of depletion calculation systems, a simple depletion calculation benchmark was performed. Values such as: -) burnup reactivity loss, -) mass balance of heavy metals and fission products (FP) were calculated. Results of the core design characteristics calculated by both JNC and Cea sides agree quite satisfactorily in terms of core conceptual design study. Potential features for improving the GCFR computational tools have been discovered during the course of this benchmark such as the way to calculate accurately the breeding gain. Different ways to improve the accuracy of the calculations have also been identified. In particular, investigation on nuclear data for steel is important for EGCR and for lumped fission products in both cores. The outcome of this benchmark is already satisfactory and will help to design more precisely GCFR cores. (authors)

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Aqueous self-cooled blanket concepts for fusion reactors

International Nuclear Information System (INIS)

Varsamis, G.; Embrechts, M.J.; Steiner, D.; Deutsch, L.; Gierszewski, P.

1987-01-01

A novel aqueous self-cooled blanket (ASCB) concept has been proposed. The water coolant also serves as the tritium breeding medium by dissolving small amounts of lithium compound in the water. The tritium recovery requirements of the ASCB concept may be facilitated by the novel in-situ radiolytic tritium separation technique in development at Chalk River Nuclear Laboratories. In this separation process deuterium gas is bubbled through the blanket coolant. Due to radiation induced processes, the equilibrium constant favors tritium migration to the deuterium gas stream. It is expected that the inherent simplicity of this design will result in a highly reliable, safe and economically attractive breeding blanket for fusion reactors. The available base of relevant information accumulated through water-cooled fission reactor programs should greatly facilitate the R and D effort required to validate the proposed blanket concept. Tests for tritium separation and corrosion compatibility show encouraging results for the feasibility of this concept

Advanced gas-cooled reactors (AGR)

Energy Technology Data Exchange (ETDEWEB)

Yeomans, R. M. [South of Scotland Electricity Board, Hunterston Power Station, West Kilbride, Ayshire, UK

1981-01-15

The paper describes the advanced gas-cooled reactor system, Hunterston ''B'' power station, which is a development of the earlier natural uranium Magnox type reactor. Data of construction, capital cost, operating performance, reactor safety and also the list of future developments are given.

Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

Energy Technology Data Exchange (ETDEWEB)

Sulaiman, S. A., E-mail: shamsulamri@tamu.edu; Dominguez-Ontiveros, E. E., E-mail: elvisdom@tamu.edu; Alhashimi, T., E-mail: jbudd123@tamu.edu; Budd, J. L., E-mail: dubaiboy@tamu.edu; Matos, M. D., E-mail: mailgoeshere@gmail.com; Hassan, Y. A., E-mail: yhasssan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, College Station, TX, 77843-3133 (United States)

2015-04-29

The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

Science.gov (United States)

Garber, Anne E.; Dickens, Ricky E.

2011-01-01

The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

Matching of dense plasma focus devices with fission reactors

International Nuclear Information System (INIS)

Harms, A.A.; Heindler, M.

1978-01-01

The potential role of dense plasma focus devices as compact neutron sources for fissile fuel breeding in conjunction with existing fission reactors is considered. It is found that advanced plasma focus devices can be used effectively in conjunction with neutronically efficient fission reactors to constitute ''self-sufficient'' breeders. Correlations among the various parameters such as the power output and conversion ratio of the fission reactor with the neutron yield and capacitor bank energy of the dense plasma focus device are presented and discussed

Comparison between two gas-cooled TRU burner subcritical reactors: fusion-fission and ADS

International Nuclear Information System (INIS)

Carluccio, T.; Rossi, P.C.R.; Angelo, G.; Maiorino, J.R.

2011-01-01

This work shows a preliminary comparative study between two gas cooled subcritical fast reactor as dedicated transuranics (TRU) transmuters: using a spallation neutron source or a D-T fusion neutron source based on ITER. The two concepts are compared in terms of a minor actinides burning performance. Further investigations are required to choose the best partition and transmutation strategy. Mainly due to geometric factors, the ADS shows better neutron multiplication. Other designs, like SABR and lead cooled ADS may show better performances than a Gas Coolead Subcritical Fast Reactors and should be investigated. We noticed that both designs can be utilized to transmutation. Besides the diverse source neutron spectra, we may notice that the geometric design and cycle parameters play a more important role. (author)

System for cooling the upper wall of a nuclear reactor vessel

International Nuclear Information System (INIS)

Pailla, Henri; Schaller, Karl; Vidard, Michel.

1974-01-01

A system for cooling the upper wall of the main vessel of a fast neutron reactor is described. This vessel is suspended from an upper shield by the upper wall. It includes coils carrying a coolant which are immersed in an intermediate liquid bathing the wall and contained in a tank integral with the vessel. At least one of the two cooling and intermediate liquids is a liquid metal. The main vessel is contained in a safety vessel, the space between the main and safety vessels is occluded in its upper part by an insulating shield placed under the tank. There is a liquid metal seal between the upper wall and the upper shield under the tank. This system has been specially designed for sodium cooled fast neutron reactors [fr

Emergency core cooling system for a fast reactor

International Nuclear Information System (INIS)

Johnson, H.G.; Madsen, R.N.

1976-01-01

The main heat transport system for a liquid-metal-cooled nuclear reactor is constructed with elevated piping and guard vessels or pipes around all components of the system below the elevation of the elevated piping so the head developed by the pumps at emergency motor speed will be unsufficient to lift the liquid-metal-coolant over the top of the guard tanks or pipes or out of the elevated piping in the event of a loss-of-coolant accident. In addition, inlet downcomers to the reactor vessel are contained within guard standpipes having a clearance volume as small as practicable. 4 claims, 2 drawing figures

Journey from discovery of nuclear fission to accelerator-driven sub-critical reactor systems (ADS)

International Nuclear Information System (INIS)

Kapoor, S.S.

2005-01-01

The epoch making discovery of nuclear fission in 1939, which resulted purely from the curiosity driven basic research to understand the atomic and nuclear structure has changed the world forever with the onset of a new era in the history of human civilization. The basic nuclear physics research pursued after the discovery of fission has also been of much relevance in the harnessing of nuclear energy. In the recent years, there is considerable interest towards developing accelerator driven sub-critical reactor systems (ADS) for the incineration of the long-lived spent fuel radioactive waste and for the utilization of thorium fuel for nuclear power generation. In this talk, we discuss important milestones in the journey from discovery of nuclear fission to ADS. (author)

Sea water take-up facility for cooling reactor auxiliary

International Nuclear Information System (INIS)

Numata, Noriko; Mizutani, Akira; Hirako, Shizuka; Uchiyama, Yuichi; Oda, Atsushi.

1997-01-01

The present invention provides an improvement of a cooling sea water take-up facility for cooling auxiliary equipments of nuclear power plant. Namely, an existent sea water take-up facility for cooling reactor auxiliary equipments has at least two circulation water systems and three independent sea water systems for cooling reactor auxiliary equipments. In this case, a communication water channel is disposed, which connects the three independent sea water systems for cooling reactor auxiliary equipments mutually by an opening/closing operation of a flow channel partitioning device. With such a constitution, even when any combination of two systems among the three circulation water systems is in inspection at the same time, one system for cooling the reactor auxiliary equipments can be kept operated, and one system is kept in a stand-by state by the communication water channel upon periodical inspection of water take-up facility for cooling the auxiliary equipments. As a result, the sea water take-up facility for cooling auxiliary equipments of the present invention have operation efficiency higher than that of a conventional case while keeping the function and safety at the same level as in the conventional case. (I.S.)

Fusion-Fission Hybrid for Fissile Fuel Production without Processing

Energy Technology Data Exchange (ETDEWEB)

Fratoni, M; Moir, R W; Kramer, K J; Latkowski, J F; Meier, W R; Powers, J J

2012-01-02

Two scenarios are typically envisioned for thorium fuel cycles: 'open' cycles based on irradiation of {sup 232}Th and fission of {sup 233}U in situ without reprocessing or 'closed' cycles based on irradiation of {sup 232}Th followed by reprocessing, and recycling of {sup 233}U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile material in a fusion-fission hybrid reactor and burning the same fuel in a critical reactor without any reprocessing or reconditioning. This fuel cycle requires the hybrid and the critical reactor to use the same fuel form. TRISO particles embedded in carbon pebbles were selected as the preferred form of fuel and an inertial laser fusion system featuring a subcritical blanket was combined with critical pebble bed reactors, either gas-cooled or liquid-salt-cooled. The hybrid reactor was modeled based on the earlier, hybrid version of the LLNL Laser Inertial Fusion Energy (LIFE1) system, whereas the critical reactors were modeled according to the Pebble Bed Modular Reactor (PBMR) and the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) design. An extensive neutronic analysis was carried out for both the hybrid and the fission reactors in order to track the fuel composition at each stage of the fuel cycle and ultimately determine the plant support ratio, which has been defined as the ratio between the thermal power generated in fission reactors and the fusion power required to breed the fissile fuel burnt in these fission reactors. It was found that the maximum attainable plant support ratio for a thorium fuel cycle that employs neither enrichment nor reprocessing is about 2. This requires tuning the neutron energy towards high energy for breeding and towards thermal energy for burning. A high fuel loading in the pebbles allows a faster spectrum in the hybrid blanket; mixing dummy carbon pebbles with fuel pebbles enables a softer spectrum in

Neutron irradiation facilities for fission and fusion reactor materials studies

International Nuclear Information System (INIS)

Rowcliffe, A.F.

1985-01-01

The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs

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.

Euratom contributions in Fast Reactor research programmes

International Nuclear Information System (INIS)

Fanghänel, Th.; Somers, J.

2013-01-01

The Sustainable Nuclear Initiative: • demonstrate long-term sustainability of nuclear energy; • demonstration reactors of Gen IV: •more efficient use of resources; • closed fuel cycle; • reduced proliferation risks; • enhanced safety features. • Systems pursued in Europe: • Sodium-cooled fast reactor SFR; • Lead-cooled fast reactor LFR; • Gas-cooled fast reactor GFR. Sustainable Nuclear Energy Technology Platform SNE-TP promotes research, development and demonstration of the nuclear fission technologies necessary to achieve the SET-Plan goals

Fission product release from fuel of water-cooled reactors

International Nuclear Information System (INIS)

Strupczewski, A.; Marks, P.; Klisinska, M.

1997-01-01

The report contains a review of theoretical models and experimental works of gaseous and volatile fission products from uranium dioxide fuel. The experimental results of activity release at low burnup and the model of fission gas behaviour at initial stage of fuel operational cycle are presented. Empirical models as well as measured results of transient fission products release rate in the temperature up to UO 2 melting point, with consideration of their chemical reactions with fuel and cladding, are collected. The theoretical and experimental data were used for calculations of gaseous and volatile fission products release, especially iodine and caesium, to the gas volume of WWER-1000 and WWER-440 type fuel rods at low and high burnup and their further release from defected rods at the assumed loss-of-coolant accident. (author)

Expected value of finite fission chain lengths of pulse reactors

International Nuclear Information System (INIS)

Liu Jianjun; Zhou Zhigao; Zhang Ben'ai

2007-01-01

The average neutron population necessary for sponsoring a persistent fission chain in a multiplying system, is discussed. In the point reactor model, the probability function θ(n, t 0 , t) of a source neutron at time t 0 leading to n neutrons at time t is dealt with. The non-linear partial differential equation for the probability generating function G(z; t 0 , t) is derived. By solving the equation, we have obtained an approximate analytic solution for a slightly prompt supercritical system. For the pulse reactor Godiva-II, the mean value of finite fission chain lengths is estimated in this work and shows that the estimated value is reasonable for the experimental analysis. (authors)

Replacement of the cooling system of the TRIGA Mainz reactor

International Nuclear Information System (INIS)

Menke, H.

1988-01-01

The inspection of the reactor facility resulted in a recommendation to install a new heat exchanger and at the same time to separate the primary cooling circuit and the water purification system. Due to possible the deposition of lime and organic matter on the tubes, the heat transfer rate has decreased. In the meantime a rule has been introduced, according to which the pressure in the secondary cooling circuit must be permanently higher than in the primary cooling circuit which prompted the design of a new cooling system. The detail planning was completed in December 1987. In response to the regulatory requirements a motion for a replacement of the cooling system was submitted to the authorities. The start of the procedure is possible a year after the obtaining of the licenses. In the planning of the changes an upgrading of the steady state power to 300 kW is envisioned

Design considerations for economically competitive sodium cooled fast reactors

International Nuclear Information System (INIS)

Zhang, Hongbin; Zhao, Haihua; Mousseau, Vincent; Szilard, Ronaldo

2009-01-01

The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phenix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design. (author)

Loss of coolant accident mitigation for liquid metal cooled space reactors

International Nuclear Information System (INIS)

Georgevich, Vladimir; Best, Frederick; Erdman, Carl

1989-01-01

A loss of coolant accident (LOCA) in a liquid metal-cooled space reactor system has been considered as a possible accident scenario. Development of new concepts that will prevent core damage by LOCA caused elevated temperatures is the primary motivation of this work. Decay heat generated by the fission products in the reactor core following shutdown is sufficiently high to melt the fuel unless energy can be removed from the pins at a sufficiently rapid rate. There are two major reasons that prevent utilization of traditional emergency cooling methods. One is the absence of gravity and the other is the vacuum condition outside the reactor vessel. A concept that overcomes both problems is the Saturated Wick Evaporation Method (SWEM). This method involves placing wicking structures at specific locations in the core to act as energy sinks. One of its properties is the isothermal behaviour of the liquid in the wick. The absorption of energy by the surface at the isothermal temperature will direct the energy into an evaporation process and not in sensible heat addition. The use of this concept enables establishment of isothermal positions within the core. A computer code that evaluates the temperature distribution of the core has been developed and the results show that this design will prevent fuel meltdown. (author)

Accident analysis of heat pipe cooled and AMTEC conversion space reactor system

International Nuclear Information System (INIS)

Yuan, Yuan; Shan, Jianqiang; Zhang, Bin; Gou, Junli; Bo, Zhang; Lu, Tianyu; Ge, Li; Yang, Zijiang

2016-01-01

Highlights: • A transient analysis code TAPIRS for HPS has been developed. • Three typical accidents are analyzed using TAPIRS. • The reactor system has the self-stabilization ability under accident conditions. - Abstract: A space power with high power density, light weight, low cost and high reliability is of crucial importance to future exploration of deep space. Space reactor is an excellent candidate because of its unique characteristics of high specific power, low cost, strong environment adaptability and so on. Among all types of space reactors, heat pipe cooled space reactor, which adopts the passive heat pipe (HP) as core cooling component, is considered as one of the most promising choices and is widely studied all over the world. This paper develops a transient analysis code (TAPIRS) for heat pipe cooled space reactor power system (HPS) based on point reactor kinetics model, lumped parameter core heat transfer model, combined HP model (self-diffusion model, flat-front startup model and network model), energy conversion model of Alkali Metal Thermal-to-Electric Conversion units (AMTEC), and HP radiator model. Three typical accidents, i.e., control drum failure, AMTEC failure and partial loss of the heat transfer area of radiator are then analyzed using TAPIRS. By comparing the simulation results of the models and steady state with those in the references, the rationality of the models and the solution method is validated. The results show the following. (1) After the failure of one set of control drums, the reactor power finally reaches a stable value after two local peaks under the temperature feedback. The fuel temperature rises rapidly, however it is still under safe limit. (2) The fuel temperature is below a safe limit under the AMTEC failure and partial loss of the heat transfer area of radiator. This demonstrates the rationality of the system design and the potential applicability of the TAPIRS code for the future engineering application of

Mirror hybrid (fusion--fission) reactor

International Nuclear Information System (INIS)

Bender, D.J.; Lee, J.D.; Neef, W.S.; Devoto, R.S.; Galloway, T.R.; Fink, J.H.; Schultz, K.R.; Culver, D.; Rao, S.

1977-10-01

The reference mirror hybrid reactor design performed by LLL and General Atomic is summarized. The reactor parameters have been chosen to minimize the cost of producing fissile fuel for consumption in fission power reactors. As in the past, we have emphasized the use of existing technology where possible and a minimum extrapolation of technology otherwise. The resulting reactor may thus be viewed as a comparatively near-term goal of the fusion program, and we project improved performance for the hybrid in the future as more advanced technology becomes available

Passive Decay Heat Removal System Options for S-CO2 Cooled Micro Modular Reactor

International Nuclear Information System (INIS)

Moon, Jangsik; Jeong, Yong Hoon; Lee, Jeong Ik

2014-01-01

To achieve modularization of whole reactor system, Micro Modular Reactor (MMR) which has been being developed in KAIST took S-CO 2 Brayton power cycle. The S-CO 2 power cycle is suitable for SMR due to high cycle efficiency, simple layout, small turbine and small heat exchanger. These characteristics of S-CO 2 power cycle enable modular reactor system and make reduced system size. The reduced size and modular system motived MMR to have mobility by large trailer. Due to minimized on-site construction by modular system, MMR can be deployed in any electricity demand, even in isolated area. To achieve the objective, fully passive safety systems of MMR were designed to have high reliability when any offsite power is unavailable. In this research, the basic concept about MMR and Passive Decay Heat Removal (PDHR) system options for MMR are presented. LOCA, LOFA, LOHS and SBO are considered as DBAs of MMR. To cope with the DBAs, passive decay heat removal system is designed. Water cooled PDHR system shows simple layout, but has CCF with reactor systems and cannot cover all DBAs. On the other hand, air cooled PDHR system with two-phase closed thermosyphon shows high reliability due to minimized CCF and is able to cope with all DBAs. Therefore, the PDHR system of MMR will follows the air-cooled PDHR system and the air cooled system will be explored

Gas-cooled reactors

International Nuclear Information System (INIS)

Schulten, R.; Trauger, D.B.

1976-01-01

Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

The story of fission reactors: from Chicago Pile to advanced energy systems

International Nuclear Information System (INIS)

Kannan, Umasankari

2017-01-01

Nuclear reactors have been designed which cater to different applications from small research reactors of a few watts to power reactors of several Giga Watts. Based on the neutron energy, there are thermal, intermediate and fast reactors operating are being designed. On the fuel utilization front, there are designs ranging from reactors using natural uranium fuel to enriched uranium to more efficient thorium based reactors. Reactors have also been designed which are neutron eaters, minor actinide burners and breeders. There have been variety of coolant and moderating materials used for different applications from water, gas cooled, liquid sodium cooled to molten salt cooled reactors. Several new reactor designs have been developed using innovative concepts in high temperature reactors, nuclear power packs and compact reactors for special purposes. The design challenges are many from modest designs to complicated hybrid reactors. The GEN-IV forum of IAEA has selected a few of these reactor designs for commercial power production in the coming years based on several quantified indicators. The evolutionary and revolutionary design approaches have been made over the years catering to different need of energy generation. A glimpse of some of the reactors being currently developed and the design modifications done in existing reactors have been given in this paper

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

International Nuclear Information System (INIS)

Li Sifeng; Li Fu; Ma Yuanle; Shi Lei

2007-01-01

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

Study on MAs transmutation of accelerator-driven system sodium-cooled fast reactor loaded with metallic fuel

International Nuclear Information System (INIS)

Han Song; Yang Yongwei

2007-01-01

Through the analysis of the effect of heavy metal actinides on the effective multiplication constant (k eff ) of the core in accelerator-driven system (ADS) sodium-cooled fast reactor loaded with metallic fuel, we gave the method for determining fuel components. the characteristics of minor actinides (MAs) transmutation was analyzed in detail. 3D burn-up code COUPLE, which couples MCNP4c3 and ORIGEN2, was applied to the neutron simulation and burn up calculation. The results of optimized scheme shows that adjusting the proportion of 239 Pu and maintaining the value during the burn-up cycle is an efficient method of designing k eff and keeping stable during the burn-up cycle. Spallation neutrons lead to the neutron spectrum harder at inner core than that at outer core. It is in favor of improving MA's fission cross sections and the capture-to-fission ratio. The total MAs transmutation support ratio 8.3 achieves excellent transmutation effect. For higher flux at inner core leads to obvious differences on transmutation efficiency,only disposing MAs at inner core is in favor of decreasing the loading mass and improving MAs transmutation effect. (authors)

Non Nuclear Testing of Reactor Systems In The Early Flight Fission Test Facilities (EFF-TF)

International Nuclear Information System (INIS)

Van Dyke, Melissa; Martin, James

2004-01-01

The Early Flight Fission-Test Facility (EFF-TF) can assist in the design and development of systems through highly effective non-nuclear testing of nuclear systems when technical issues associated with near-term space fission systems are 'non-nuclear' in nature (e.g. system's nuclear operations are understood). For many systems, thermal simulators can be used to closely mimic fission heat deposition. Axial power profile, radial power profile, and fuel pin thermal conductivity can be matched. In addition to component and subsystem testing, operational and lifetime issues associated with the steady state and transient performance of the integrated reactor module can be investigated. Instrumentation at the EFF-TF allows accurate measurement of temperature, pressure, strain, and bulk core deformation (useful for accurately simulating nuclear behavior). Ongoing research at the EFF-TF is geared towards facilitating research, development, system integration, and system utilization via cooperative efforts with DOE laboratories, industry, universities, and other Nasa centers. This paper describes the current efforts for the latter portion of 2003 and beginning of 2004. (authors)

Gas cooled reactors

International Nuclear Information System (INIS)

Kojima, Masayuki.

1985-01-01

Purpose: To enable direct cooling of reactor cores thereby improving the cooling efficiency upon accidents. Constitution: A plurality sets of heat exchange pipe groups are disposed around the reactor core, which are connected by way of communication pipes with a feedwater recycling device comprising gas/liquid separation device, recycling pump, feedwater pump and emergency water tank. Upon occurrence of loss of primary coolants accidents, the heat exchange pipe groups directly absorb the heat from the reactor core through radiation and convection. Although the water in the heat exchange pipe groups are boiled to evaporate if the forcive circulation is interrupted by the loss of electric power source, water in the emergency tank is supplied due to the head to the heat exchange pipe groups to continue the cooling. Furthermore, since the heat exchange pipe groups surround the entire circumference of the reactor core, cooling is carried out uniformly without resulting deformation or stresses due to the thermal imbalance. (Sekiya, K.)

Fission product release from SLOWPOKE-2 reactors

Energy Technology Data Exchange (ETDEWEB)

Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

1994-12-31

Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

Monte Carlo modeling of Lead-Cooled Fast Reactor in adiabatic equilibrium state

Energy Technology Data Exchange (ETDEWEB)

Stanisz, Przemysław, E-mail: pstanisz@agh.edu.pl; Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

2016-05-15

Graphical abstract: - Highlights: • We present the Monte Carlo modeling of the LFR in the adiabatic equilibrium state. • We assess the adiabatic equilibrium fuel composition using the MCB code. • We define the self-adjusting process of breeding gain by the control rod operation. • The designed LFR can work in the adiabatic cycle with zero fuel breeding. - Abstract: Nuclear power would appear to be the only energy source able to satisfy the global energy demand while also achieving a significant reduction of greenhouse gas emissions. Moreover, it can provide a stable and secure source of electricity, and plays an important role in many European countries. However, nuclear power generation from its birth has been doomed by the legacy of radioactive nuclear waste. In addition, the looming decrease in the available resources of fissile U235 may influence the future sustainability of nuclear energy. The integrated solution to both problems is not trivial, and postulates the introduction of a closed-fuel cycle strategy based on breeder reactors. The perfect choice of a novel reactor system fulfilling both requirements is the Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state. In such a state, the reactor converts depleted or natural uranium into plutonium while consuming any self-generated minor actinides and transferring only fission products as waste. We present the preliminary design of a Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state with the Monte Carlo Continuous Energy Burnup Code – MCB. As a reference reactor model we apply the core design developed initially under the framework of the European Lead-cooled SYstem (ELSY) project and refined in the follow-up Lead-cooled European Advanced DEmonstration Reactor (LEADER) project. The major objective of the study is to show to what extent the constraints of the adiabatic cycle are maintained and to indicate the phase space for further improvements. The analysis

Reactor core cooling device

International Nuclear Information System (INIS)

Kobayashi, Masahiro.

1986-01-01

Purpose: To safely and effectively cool down the reactor core after it has been shut down but is still hot due to after-heat. Constitution: Since the coolant extraction nozzle is situated at a location higher than the coolant injection nozzle, the coolant sprayed from the nozzle, is free from sucking immediately from the extraction nozzle and is therefore used effectively to cool the reactor core. As all the portions from the top to the bottom of the reactor are cooled simultaneously, the efficiency of the reactor cooling process is increased. Since the coolant extraction nozzle can be installed at a point considerably higher than the coolant injection nozzle, the distance from the coolant surface to the point of the coolant extraction nozzle can be made large, preventing cavitation near the coolant extraction nozzle. Therefore, without increasing the capacity of the heat exchanger, the reactor can be cooled down after a shutdown safely and efficiently. (Kawakami, Y.)

A study of silver behavior in Gas-turbine High Temperature Gas-cooled Reactor

International Nuclear Information System (INIS)

Sawa, Kazuhiro; Tanaka, Toshiyuki

1995-11-01

A Gas-turbine High Temperature Gas-cooled Reactor (GT-HTGR) is one of the promising reactor systems of future HTGRs. In the design of GT-HTGR, behavior of fission products, especially of silver, is considered to be important from the view point of maintenance of gas-turbine. A study of silver behavior in the GT-HTGR was carried out based on current knowledge. The purposes of this study were to determine an importance of the silver problem quantitatively, countermeasures to the problem and items of future research and development which will be needed. In this study, inventory, fractional release from fuel, plateout in the primary circuit and radiation dose were evaluated, respectively. Based on this study, it is predicted that gamma-ray from plateout silver in gas-turbine system contributes about a half of total radiation dose after reactor shutdown. In future, more detail data for silver release from fuel, plateout behavior, etc. using the High Temperature Engineering Test Reactor (HTTR), for example, will be needed to carry out reasonable design. (author)

Contained fissionly vaporized imploded fission explosive breeder reactor

International Nuclear Information System (INIS)

Marwick, E.F.

1978-01-01

Disclosed is a nuclear reactor system which produces useful thermal power and breeds fissile isotopes wherein large spherical complex slugs containing fissile and fertile isotopes as well as vaporizing and tamping materials are exploded seriatim in a large containing chamber having walls protected from the effects of the explosion by about two thousand tons of slurry of fissile and fertile isotopes in molten alkali metal. The slug which is slightly sub-critical prior to its entry into the centroid portion of the chamber, then becomes slightly more than prompt-critical because of the near proximity of neutron-reflecting atoms and of fissioning atoms within the slurry. The slurry is heated by explosion of the slugs and serves as a working fluid for extraction of heat energy from the reactor. Explosive debris is precipitated from the slurry and used for the fabrication of new slugs

Material Issues of Blanket Systems for Fusion Reactors - Compatibility with Cooling Water -

Science.gov (United States)

Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro

Environmental assisted cracking (EAC) is one of the material issues for the reactor core components of light water power reactors(LWRs). Much experience and knowledge have been obtained about the EAC in the LWR field. They will be useful to prevent the EAC of water-cooled blanket systems of fusion reactors. For the austenitic stainless steels and the reduced-activation ferritic/martensitic steels, they clarifies that the EAC in a water-cooled blanket does not seem to be acritical issue. However, some uncertainties about influences on water temperatures, water chemistries and stress conditions may affect on the EAC. Considerations and further investigations elucidating the uncertainties are discussed.

Cleaning device for recycling pump motor cooling system in nuclear reactor

International Nuclear Information System (INIS)

Katayama, Kenjiro; Kondo, Takahisa; Shindo, Kenjiro; Akimoto, Jun.

1996-01-01

The cleaning device of the present invention comprises a cleaning water supply pump, a filter for filtering the cleaning water and a cap member for isolating the inside of a motor casing from the inside of a reactor pressure vessel. A motor in the motor casing and a pump in the reactor pressure vessel are removed, the cap member is attached to the upper end of the motor casing to isolate the inside of the motor casing from the inside of the reactor pressure vessel. If the cleaning water supply pump is operated in this state, the cleaning water flows from a returning pipeline for cooling water circulation, connected to the motor casing to supply pipelines through a heat exchange and is discharged. The discharged water passes through a filter and is sent again, as the cleaning water, to the cleaning water supply pump. With such procedures, the recycling pump motor cooling system in the BWR type reactor can be cleaned without disposing a cyclone separator and irrespective of presence or absence of reactor coolants in the reactor pressure vessel. (I.N.)

Emergency core cooling system

International Nuclear Information System (INIS)

Ando, Masaki.

1987-01-01

Purpose: To actuate an automatic pressure down system (ADS) and a low pressure emergency core cooling system (ECCS) upon water level reduction of a nuclear reactor other than loss of coolant accidents (LOCA). Constitution: ADS in a BWR type reactor is disposed for reducing the pressure in a reactor container thereby enabling coolant injection from a low pressure ECCS upon LOCA. That is, ADS has been actuated by AND signal for a reactor water level low signal and a dry well pressure high signal. In the present invention, ADS can be actuated further also by AND signal of the reactor water level low signal, the high pressure ECCS and not-operation signal of reactor isolation cooling system. In such an emergency core cooling system thus constituted, ADS operates in the same manner as usual upon LOCA and, further, ADS is operated also upon loss of feedwater accident in the reactor pressure vessel in the case where there is a necessity for actuating the low pressure ECCS, although other high pressure ECCS and reactor isolation cooling system are not operated. Accordingly, it is possible to improve the reliability upon reactor core accident and mitigate the operator burden. (Horiuchi, T.)

Feasibility study of self sustaining capability on water cooled thorium reactors for different power reactors

International Nuclear Information System (INIS)

Permana, S.; Takaki, N.; Sekimoto, H.

2007-01-01

Thorium fuel cycle can maintain the sustainable system of the reactor for self sustaining system for future sustainable development in the world. Some characteristics of thorium cycle show some advantages in relation to higher breeding capability, higher performance of burn-up and more proliferation resistant. Several investigations was performed to improve the breeding capability which is essential for maintaining the fissile sustainability during reactor operation in thermal reactor such as Shippingport reactor and molten salt breeder reactor (MSBR) project. The preliminary study of breeding capability on water cooled thorium reactor has been investigated for various power output. The iterative calculation system is employed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000. In this calculation, 1238 fission products and 129 heavy nuclides are employed. In the cell calculation, 26 heavy metals and 66 fission products and 1 pseudo FP are employed. The employed nuclear data library was JENDL 3.2. The reactor is fueled by 2 33U-Th Oxide and it has used the light water coolant as moderator. Some characteristics such as conversion ratio and void reactivity coefficient performances are evaluated for the systems. The moderator to fuel ratio (MFR) values and average burnups are studied for survey parameter. The parametric survey for different power outputs are employed from 10 MWt to 3000 MWt for evaluating the some characteristics of core size and leakage effects to the spectra profile, required enrichment, breeding capability, fissile inventory condition, and void reactivity coefficient. Different power outputs are employed in order to evaluate its effect to the required enrichment for criticality, breeding capability, void reactivity and fissile inventory accumulation. The obtained value of the conversion ratios is evaluated by using the equilibrium atom composition. The conversion ratio is employed based on the

Cooling system for the connecting rings of a fast neutron reactor vessel

International Nuclear Information System (INIS)

Martin, J.-P.; Malaval, Claude

1974-01-01

A description is given of a cooling system for the vessel connecting rings of a fast neutron nuclear reactor, particularly of a main vessel containing the core of the reactor and a volume of liquid metal coolant at high temperature and a safety vessel around the main vessel, both vessels being suspended to a rigid upper slab kept at a lower temperature. It is mounted in the annular space between the two vessels and includes a neutral gas circuit set up between the wall of the main vessel to be cooled and that of the safety vessel itself cooled from outer. The neutral gas system comprises a plurality of ventilators fitted in holes made through the thickness of the upper slab and opening on to the space between the two vessels. It also includes two envelopes lining the walls of these vessels, establishing with them small section channels for the circulation of the neutral gas cooled against the safety vessel and heated against the main vessel [fr

Behaviour of gas cooled reactor fuel under accident conditions

International Nuclear Information System (INIS)

1991-11-01

The Specialists Meeting on Behaviour of Gas Cooled Reactor Fuel under Accident Conditions was convened by the International Atomic Energy Agency on the recommendation of the International Working Group on Gas Cooled Reactors. The purpose of the meeting was to provide an international forum for the review of the development status and for the discussion on the behaviour of gas cooled reactor fuel under accident conditions and to identify areas in which additional research and development are still needed and where international co-operation would be beneficial for all involved parties. The meeting was attended by 45 participants from France, Germany, Japan, Switzerland, the Union of Soviet Socialists Republics, the United Kingdom, the United States of America, CEC and the IAEA. The meeting was subdivided into five technical sessions: Summary of Current Research and Development Programmes for Fuel; Fuel Manufacture and Quality Control; Safety Requirements; Modelling of Fission Product Release - Part I and Part II; Irradiation Testing/Operational Experience with Fuel Elements; Behaviour at Depressurization, Core Heat-up, Power Transients; Water/Steam Ingress - Part I and Part II. 22 papers were presented. A separate abstract was prepared for each of these papers. At the end of the meeting a round table discussion was held on Directions for Future R and D Work and International Co-operation. Refs, figs and tabs

Advanced burnup calculation code system in a subcritical state with continuous-energy Monte Carlo code for fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Matsunaka, Masayuki; Ohta, Masayuki; Miyamaru, Hiroyuki; Murata, Isao

2009-01-01

The fusion-fission (FF) hybrid reactor is a promising energy source that is thought to act as a bridge between the existing fission reactor and the genuine fusion reactor in the future. The burnup calculation system that aims at precise burnup calculations of a subcritical system was developed for the detailed design of the FF hybrid reactor, and the system consists of MCNP, ORIGEN, and postprocess codes. In the present study, the calculation system was substantially modified to improve the calculation accuracy and at the same time the calculation speed as well. The reaction rate estimation can be carried out accurately with the present system that uses track-length (TL) data in the continuous-energy treatment. As for the speed-up of the reaction rate calculation, a new TL data bunching scheme was developed so that only necessary TL data are used as long as the accuracy of the point-wise nuclear data is conserved. With the present system, an example analysis result for our proposed FF hybrid reactor is described, showing that the computation time could really be saved with the same accuracy as before. (author)

Fusion-fission hybrid reactors

International Nuclear Information System (INIS)

Greenspan, E.

1984-01-01

This chapter discusses the range of characteristics attainable from hybrid reactor blankets; blanket design considerations; hybrid reactor designs; alternative fuel hybrid reactors; multi-purpose hybrid reactors; and hybrid reactors and the energy economy. Hybrid reactors are driven by a fusion neutron source and include fertile and/or fissile material. The fusion component provides a copious source of fusion neutrons which interact with a subcritical fission component located adjacent to the plasma or pellet chamber. Fissile fuel and/or energy are the main products of hybrid reactors. Topics include high F/M blankets, the fissile (and tritium) breeding ratio, effects of composition on blanket properties, geometrical considerations, power density and first wall loading, variations of blanket properties with irradiation, thermal-hydraulic and mechanical design considerations, safety considerations, tokamak hybrid reactors, tandem-mirror hybrid reactors, inertial confinement hybrid reactors, fusion neutron sources, fissile-fuel and energy production ability, simultaneous production of combustible and fissile fuels, fusion reactors for waste transmutation and fissile breeding, nuclear pumped laser hybrid reactors, Hybrid Fuel Factories (HFFs), and scenarios for hybrid contribution. The appendix offers hybrid reactor fundamentals. Numerous references are provided

Pressure due to fission gases in a fuel element circulating in a reactor

International Nuclear Information System (INIS)

Fonteray, Jean

1965-01-01

This document states calculation hypotheses and methods used to assess pressures due to fission gases in a fuel element moving in a reactor channel in the reverse direction with respect to the cooling fluid. The calculation comprises the calculation of the temperature in the fuel rod, of the reduced diffusion coefficient, of the diffused gas fraction, of the pressure. The appendix describes the use of the SPM 076 software: input data, output results, computing time [fr

Estimation on the Pressure Loss of the Conceptual Primary Cooling System and Design of the Primary Cooling Pump for a Research Reactor

International Nuclear Information System (INIS)

Seo, Kyoung Woo; Oh, Jae Min; Park, Jong Hark; Chae, Hee Taek; Seo, Jae Kwang; Park, Cheon Tae; Yoon, Ju Hyeon; Lee, Doo Jeong

2009-01-01

A new conceptual primary cooling system (PCS) for a research reactor has been designed for an adequate cooling to the reactor core which has various powers ranging from 30MW through 80MW. The developed primary cooling system consisted of decay tanks, pumps, heat exchangers, vacuum breakers, some isolation and check valves, connection piping, and instruments. Because the system flow rate should be determined by the thermal hydraulic design analysis for the core, the heads to design the primary cooling pumps (PCPs) in a PCS will be estimated by the variable system flow rates. The heads of the part of a research reactor vessel was evaluated by the previous study. The various pressure losses of the PCS can be calculated by the dimensional analysis of the pipe flow and the head loss coefficient of the components. The purpose of this research is to estimate the various pressure losses and to design the PCPs

78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

Science.gov (United States)

2013-10-25

... comments were received. A companion guide, DG-1277, ``Initial Test Program of Emergency Core Cooling... NUCLEAR REGULATORY COMMISSION [NRC-2011-0129] Preoperational Testing of Emergency Core Cooling... (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors...

Investigation/evaluation of water cooled fast reactor in the feasibility study on commercialized fast reactor cycle systems. Intermediate evaluation of phase-II study

International Nuclear Information System (INIS)

Kotake, Syoji; Nishikawa, Akira

2005-01-01

Feasibility study on commercialized fast reactor cycle systems aims at investigation and evaluation of FBR design requirement's attainability, operation and maintenance, and technical feasibility of the candidate system. Development targets are 1) ensuring safety, 2) economic competitiveness, 3) efficient utilization of resources, 4) reduction of environmental load and 5) enhancement of nuclear non-proliferation. Based on the selection of the promising concepts in the first phase, conceptual design for the plant system has proceeded with the following plant system: a) sodium cooled reactors at large size and medium size module reactors, b) a lead-bismuth cooled medium size reactor, c) a helium gas cooled large size reactor and d) a BWR type large size FBR. Technical development and feasibility has been assessed and the study considers the need of respective key technology development for the confirmation of the feasibility study. (T. Tanaka)

Apparatus for measuring the release of fission gases and other fission products by degassing

Energy Technology Data Exchange (ETDEWEB)

Stradal, Karl Alfred

1970-10-15

In gas-cooled high-temperature reactors, the fuel is, in general, inserted in the fuel elements in the form of small particles, which are, for example, coated with pyrolytic carbon. The purpose of this coating is to keep the fission products separate from the coolant gas. The further development of these coated particles makes it necessary to check the retention capacity. One possible method of doing this is the degassing test after irradiation in the reactor. An apparatus is described below, which was developed and installed in order to measure to a higher degree of sensitivity and in serial measurements the release of fission gases and sparingly volatile fission products.

Convective cooling in a pool-type research reactor

Energy Technology Data Exchange (ETDEWEB)

Sipaun, Susan, E-mail: susan@nm.gov.my [Malaysian Nuclear Agency, Industrial Technology Division, Blok 29T, Bangi 43200, Selangor (Malaysia); Usman, Shoaib, E-mail: usmans@mst.edu [Missouri University of Science and Technology, Nuclear Engineering, 222 Fulton Hall 301 W.14th St., Rolla 64509 MO (United States)

2016-01-22

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

Neutronics investigation of advanced self-cooled liquid blanket systems in helical reactor

International Nuclear Information System (INIS)

Tanaka, T.; Sagara, A.; Muroga, T.; Youssef, M.Z.

2006-10-01

Neutronics performances of advanced self-cooled liquid blanket systems have been investigated in design activity of the helical-type reactor FFHR2. In the present study, a new three-dimensional (3-D) neutronics calculation system has been developed for the helical-type reactor to enhance quick feedback between neutronics evaluation and design modification. Using this new calculation system, advanced Flibe-cooled and Li-cooled liquid blanket systems proposed for FFHR2 have been evaluated to make clear design issues to enhance neutronics performance. Based on calculated results, modification of the blanket dimensions and configuration have been attempted to achieve the adequate tritium breeding ability and neutron shielding performance in the helical reactor. The total tritium breeding ratios (TBRs) obtained after modifying the blanket dimensions indicated that all the advanced blanket systems proposed for FFHR2 would achieve adequate tritium self-sufficiency by dimension adjustment and optimization of structures in the breeder layers. Issues in neutron shielding performance have been investigated quantitatively using 3-D geometry of the helical blanket system, support structures, poloidal coils etc. Shielding performance of the helical coils against direct neutrons from core plasma would achieve design target by further optimization of shielding materials. However, suppression of the neutron streaming and reflection through the divertor pumping areas in the original design is important issue to protect the poloidal coils and helical coils, respectively. Investigation of the neutron wall loading indicated that the peaking factor of the neutron wall load distribution would be moderated by the toroidal and helical effect of the plasma distribution in the helical reactor. (author)

Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

International Nuclear Information System (INIS)

1975-12-01

The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

Thermodynamic data for selected gas impurities in the primary coolant of high-temperature gas-cooled reactors

International Nuclear Information System (INIS)

Feber, R.C.

1976-12-01

The literature of thermodynamic data for selected fission-product species is reviewed and supplemented in support of complex chemical equilibrium calculations applied to fission-product distributions in the primary coolant of high-temperature gas-cooled reactors. Thermodynamic functions and heats and free energies of formation are calculated and tabulated to 3000 0 K for CsI (s,l,g), Cs 2 I 2 (g), CH 3 I(g), COI 2 (g), and CsH(g). 79 references

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

International Nuclear Information System (INIS)

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

1975-01-01

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

Phase 1 space fission propulsion system design considerations

International Nuclear Information System (INIS)

Houts, Mike; Van Dyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Salvail, Pat; Hrbud, Ivana; Carter, Robert

2002-01-01

Fission technology can enable rapid, affordable access to any point in the solar system. If fission propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and operated. Studies conducted in fiscal year 2001 (IISTP, 2001) show that fission electric propulsion (FEP) systems operating at 80 kWe or above could enhance or enable numerous robotic outer solar system missions of interest. At these power levels it is possible to develop safe, affordable systems that meet mission performance requirements. In selecting the system design to pursue, seven evaluation criteria were identified: safety, reliability, testability, specific mass, cost, schedule, and programmatic risk. A top-level comparison of three potential concepts was performed: an SP-100 based pumped liquid lithium system, a direct gas cooled system, and a heatpipe cooled system. For power levels up to at least 500 kWt (enabling electric power levels of 125-175 kWe, given 25-35% power conversion efficiency) the heatpipe system has advantages related to several criteria and is competitive with respect to all. Hardware-based research and development has further increased confidence in the heatpipe approach. Successful development and utilization of a 'Phase 1' fission electric propulsion system will enable advanced Phase 2 and Phase 3 systems capable of providing rapid, affordable access to any point in the solar system

The passive safety characteristics of modular high temperature gas-cooled reactor fuel elements

International Nuclear Information System (INIS)

Goodin, D.T.; Kania, M.J.; Nabielek, H.; Schenk, W.; Verfondern, K.

1988-01-01

High-Temperature Gas-Cooled Reactors (HTGR) in both the US and West Germany use an all-ceramic, coated fuel particle to retain fission products. Data from irradiation, postirradiation examinations and postirradiation heating experiments are used to study the performance capabilities of the fuel particles. The experimental results from fission product release tests with HTGR fuel are discussed. These data are used for development of predictive fuel performance models for purposes of design, licensing, and risk analyses. During off normal events, where temperatures may reach up to 1600/degree/C, the data show that no significant radionuclide releases from the fuel will occur

Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

Science.gov (United States)

Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

2017-07-01

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

Studies on the properties of hard-spectrum, actinide fissioning reactors. Final report

International Nuclear Information System (INIS)

Nelson, J.B.; Prichard, A.W.; Schofield, P.E.; Robinson, A.H.; Spinrad, B.I.

1980-01-01

It is technically feasible to construct an operable (e.g., safe and stable) reactor to burn waste actinides rapidly. The heart of the concept is a driver core of EBR-II type, with a central radial target zone in which fuel elements, made entirely of waste actinides are exposed. This target fuel undergoes fission, as a result of which actinides are rapidly destroyed. Although the same result could be achieved in more conventionally designed LWR or LMFBR systems, the fast spectrum reactor does a much more efficient job, by virtue of the fact that in both LWR and LMFBR reactors, actinide fission is preceded by several captures before a fissile nuclide is formed. In the fast spectrum reactor that is called ABR (actinide burning reactor), these neutron captures are short-circuited

Nuclear Power from Fission Reactors. An Introduction.

Science.gov (United States)

Department of Energy, Washington, DC. Technical Information Center.

The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

Development of a Neutron Flux Monitoring System for Sodium-cooled Fast Reactors

OpenAIRE

Verma, Vasudha

2017-01-01

Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility stud...

CEA programme on gas cooled reactors

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Chapelot, Ph.; Gauthier, J.C.

2002-01-01

Future nuclear energy systems studies conducted by the CEA aim at investigating and developing promising technologies for future reactors, fuels and fuel cycles, for nuclear power to play a major part in sustainable energy policies. Reactors and fuel cycles are considered as integral parts of a nuclear system to be optimised as a whole. Major goals assigned to future nuclear energy systems are the following: reinforced economic competitiveness with other electricity generation means, with a special emphasis on reducing the investment cost; enhanced reliability and safety, through an improved management of reactor operation in normal and abnormal plant conditions; minimum production of long lived radioactive waste; resource saving through an effective and flexible use of the available resources of fissile and fertile materials; enhanced resistance to proliferation risks. The three latter goals are essential for the sustainability of nuclear energy in the long term. Additional considerations such as the potentialities for other applications than electricity generation (co-generation, production of hydrogen, sea water desalination) take on an increasing importance. Sustainability goals call for fast neutron spectra (to transmute nuclear waste and to breed fertile fuel) and for recycling actinides from the spent fuel (plutonium and minor actinides). New applications and economic competitiveness call for high temperature technologies (850 deg C), that afford high conversion efficiencies and hence less radioactive waste production and discharged heat. These orientations call for breakthroughs beyond light water reactors. Therefore, as a result of a screening review of candidate technologies, the CEA has selected an innovative concept of high temperature gas cooled reactor with a fast neutron spectrum, robust refractory fuel, direct conversion with a gas turbine, and integrated on-site fuel cycle as a promising system for a sustainable energy development. This objective

Space Fission Reactor Structural Materials: Choices Past, Present and Future

International Nuclear Information System (INIS)

Busby, Jeremy T.; Leonard, Keith J.

2007-01-01

Nuclear powered spacecraft will enable missions well beyond the capabilities of current chemical, radioisotope thermal generator and solar technologies. The use of fission reactors for space applications has been considered for over 50 years, although, structural material performance has often limited the potential performance of space reactors. Space fission reactors are an extremely harsh environment for structural materials with high temperatures, high neutron fields, potential contact with liquid metals, and the need for up to 15-20 year reliability with no inspection or preventative maintenance. Many different materials have been proposed as structural materials. While all materials meet many of the requirements for space reactor service, none satisfy all of them. However, continued development and testing may resolve these issues and provide qualified materials for space fission reactors.

Recycling option search for a 600 MWE sodium-cooled transmutation fast reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Kyo; Kim, Myung Hyun [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)

2015-02-15

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro- SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to 20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

Directory of Open Access Journals (Sweden)

Yong Kyo Lee

2015-02-01

Full Text Available Four recycling scenarios involving pyroprocessing of spent fuel (SF have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR, KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs. If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

Tritium chemistry in fission and fusion reactors

International Nuclear Information System (INIS)

Roth, E.; Masson, M.; Briec, M.

1986-09-01

We are interested in the behaviour of tritium inside the solids where it is generated both in the case of fission nuclear reactor fuel elements, and in that of blankets of future fusion reactor. In the first case it is desirable to be able to predict whether tritium will be found in the hulls or in the uranium oxide, and under what chemical form, in order to take appropriate steps for it's removal in reprocessing plants. In fusion reactors breeding large amounts of tritium and burning it in the plasma should be accomplished in as short a cycle as possible in order to limit inventories that are associated with huge activities. Mastering the chemistry of every step is therefore essential. Amounts generated are not of the same order of magnitude in the two cases studied. Ternary fissions produce about 66 10 13 Bq (18 000 Ci) per year of tritium in a 1000 MWe fission generator, i.e., about 1.8 10 10 Bq (0.5 Ci) per day per ton of fuel

Production of energy in a high temperature reactor

International Nuclear Information System (INIS)

Anon.

1979-01-01

The cooling gas having left the reactor core is fed to a generator for direct production of current from the kinetic energy. Afterwards the gas is fed to a heat exchanger for cooling, then compressed and refed to the reactor core. The method further comprises that one part of the energy of the fission material is directly converted to electric energy in the reactor core, whereas the other part of the energy of the fission material is impressed upon the cooling gas. According to the invention the cooling gas when entering the reactor is first fed to that part of the reactor core which serves as a thermoionic or thermoelectric transducer. Afterwards the cooling gas is fed to the remaining part of the reactor gas. (P.K.)

Vibration test on KMRR reactor structure and primary cooling system piping

Energy Technology Data Exchange (ETDEWEB)

Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-10-01

Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

Vibration test on KMRR reactor structure and primary cooling system piping

International Nuclear Information System (INIS)

Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo

1994-10-01

Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author)

Contained fission explosion breeder reactor system

International Nuclear Information System (INIS)

Juhl, N.H.; Marwick, E.F.

1983-01-01

A reactor system for producing useful thermal energy and valuable isotopes, such as plutonium-239, uranium-233, and/or tritium, in which a pair of sub-critical masses of fissile and fertile actinide slugs are propelled into an ellipsoidal pressure vessel. The propelled slugs intercept near the center of the chamber where the concurring slugs become a more than prompt configuration thereby producing a fission explosion. Re-useable accelerating mechanisms are provided external of the vessel for propelling the slugs at predetermined time intervals into the vessel. A working fluid of lean molten metal slurry is injected into the chamber prior to each explosion for the attenuation of the explosion's effects, for the protection of the chamber's walls, and for the absorbtion of thermal energy and debris from the explosion. The working fluid is injected into the chamber in a pattern so as not to interfere with the flight paths of the slugs and to maximize the concentration of working fluid near the chamber's center. The heated working fluid is drained from the vessel and is used to perform useful work. Most of the debris from the explosion is collected as precipitate and is used for the manufacture of new slugs

Analysis on small long life reactor using thorium fuel for water cooled and metal cooled reactor types

International Nuclear Information System (INIS)

Permana, Sidik

2009-01-01

Long-life reactor operation can be adopted for some special purposes which have been proposed by IAEA as the small and medium reactor (SMR) program. Thermal reactor and fast reactor types can be used for SMR and in addition to that program the utilization of thorium fuel as one of the candidate as a 'partner' fuel with uranium fuel which can be considered for optimizing the nuclear fuel utilization as well as recycling spent fuel. Fissile U-233 as the main fissile material for thorium fuel shows higher eta-value for wider energy range compared with other fissile materials of U-235 and Pu-239. However, it less than Pu-239 for fast energy region, but it still shows high eta-value. This eta-value gives the reactor has higher capability for obtaining breeding condition or high conversion capability. In the present study, the comparative analysis on small long life reactor fueled by thorium for different reactor types (water cooled and metal cooled reactor types). Light water and heavy water have been used as representative of water-cooled reactor types, and for liquid metal-cooled reactor types, sodium-cooled and lead-bismuth-cooled have been adopted. Core blanket arrangement as general design configuration, has been adopted which consist of inner blanket region fueled by thorium oxide, and two core regions (inner and out regions) fueled by fissile U-233 and thorium oxide with different percentages of fissile content. SRAC-CITATION and JENDL-33 have been used as core optimization analysis and nuclear data library for this analysis. Reactor operation time can reaches more than 10 years operation without refueling and shuffling for different reactor types and several power outputs. As can be expected, liquid metal cooled reactor types can be used more effective for obtaining long life reactor with higher burnup, higher power density, higher breeding capability and lower excess reactivity compared with water-cooled reactors. Water cooled obtains long life core operation

Primary system fission product release and transport. A state-of-the-art report to the committee on the safety of nuclear installations

International Nuclear Information System (INIS)

Wright, A.L.

1994-06-01

This report presents a summary of the status of research activities associated with fission product behavior (release and transport) under severe accident conditions within the primary systems of water-moderated and water-cooled nuclear reactors. For each of the areas of fission product release and fission product transport, the report summarizes relevant information on important phenomena, major experiments performed, relevant computer models and codes, comparisons of computer code calculations with experimental results, and general conclusions on the overall state of the art. Finally, the report provides an assessment of the overall importance and knowledge of primary system release and transport phenomena and presents major conclusions on the state of the art

Primary system fission product release and transport: A state-of-the-art report to the committee on the safety of nuclear installations

International Nuclear Information System (INIS)

Wright, A.L.

1994-06-01

This report presents a summary of the status of research activities associated with fission product behavior (release and transport) under severe accident conditions within the primary systems of water-moderated and water-cooled nuclear reactors. For each of the areas of fission product release and fission product transport, the report summarizes relevant information on important phenomena, major experiments performed, relevant computer models and codes, comparisons of computer code calculations with experimental results, and general conclusions on the overall state of the art. Finally, the report provides an assessment of the overall importance and knowledge of primary system release and transport phenomena and presents major conclusions on the state of the art

Primary system fission product release and transport: A state-of-the-art report to the committee on the safety of nuclear installations

Energy Technology Data Exchange (ETDEWEB)

Wright, A.L. [Oak Ridge National Lab., TN (United States)

1994-06-01

This report presents a summary of the status of research activities associated with fission product behavior (release and transport) under severe accident conditions within the primary systems of water-moderated and water-cooled nuclear reactors. For each of the areas of fission product release and fission product transport, the report summarizes relevant information on important phenomena, major experiments performed, relevant computer models and codes, comparisons of computer code calculations with experimental results, and general conclusions on the overall state of the art. Finally, the report provides an assessment of the overall importance and knowledge of primary system release and transport phenomena and presents major conclusions on the state of the art.

Power distribution monitoring system in the boiling water cooled reactor core

International Nuclear Information System (INIS)

Leshchenko, Yu.I.; Sadulin, V.P.; Semidotskij, I.I.

1987-01-01

Consideration is being given to the system of physical power distribution monitoring, used during several years in the VK-50 tank type boiling water cooled reactor. Experiments were conducted to measure the ratios of detector prompt and activation currents, coefficients of detector relative sensitivity with respect to neutrons and effective cross sections of 103 Rh interaction with thermal and epithermal neutrons. Mobile self-powered detectors (SPD) with rhodium emitters are used as the power distribution detectors in the considered system. All detectors move simultaneously with constant rate in channels, located in fuel assembly central tubes, when conducting the measurements. It is concluded on the basis of analyzing the obtained data, that investigated system with calibrated SPD enables to monitor the absolute power distribution in fuel assemblies under conditions of boiling water cooled reactor and is independent of thermal engineering measurements conducted by in core instruments

Utilization of fast reactor excess neutrons for burning long-lived fission products

International Nuclear Information System (INIS)

Kawashima, K.; Kobayashi, K.; Kaneto, K.

1995-01-01

An evaluation is made on a large MOX fuel fast reactor's capability of burning long lived fission product Tc-99, which dominates the long term radiotoxicity of the high level radioactive waste. The excess neutrons generated in the fast reactor core are utilized to transmute Tc-99 to stable isotopes due to neutron capture reaction. The fission product target assemblies which consist of Tc-99 are charged to the reactor core periphery. The fission product target neutrons are moderated to a great deal to pursue the possibility of enhancing the transmutation rate. Any impacts of loading the fission product target assemblies on the core nuclear performances are assessed. A long term Tc-99 accumulation scenario is considered in the mix of fission product burner fast reactor and non-burner LWRs. (author)

Unlimited cooling capacity of the passive-type emergency core cooling system of the MARS reactor

International Nuclear Information System (INIS)

Bandini, G.; Caira, M.; Naviglio, A.; Sorabella, L.

1995-01-01

The MARS nuclear plant is equipped with a 600 MWth PWR type nuclear steam supply system, with completely innovative engineered core safeguards. The most relevant innovative safety system of this plant is its Emergency Core Cooling System, which is completely passive (with only one non static component). The Emergency Core Cooling System (ECCS) of the MARS reactor is natural-circulation, passive-type, and its intervention follows a core flow decrease, whatever was the cause. The operation of the system is based on a cascade of three fluid systems, functionally interfacing through heat exchangers; the first fluid system is connected to the reactor vessel and the last one includes an atmospheric-pressure condenser, cooled by external air. The infinite thermal capacity of the final heat sink provides the system an unlimited autonomy. The capability and operability of the system are based on its integrity and on the integrity of the primary coolant boundary (both of them are permanently enclosed in a pressurized containment; 100% redundancy is also foreseen) and on the operation of only one non static component (a check valve), with 400% redundancy. In the paper, all main thermal hydraulic transients occurring as a consequence of postulated accidents are analysed, to verify the capability of the passive-type ECCS to intervene always in time, without causing undue conditions of reduced coolability of the core (DNB, etc.), and to verify its capability to guarantee a long-term (indefinite) coolability of the core without the need of any external intervention. (author)

Emergency cooling system for nuclear reactors

International Nuclear Information System (INIS)

Frisch, E.; Andrews, H.N.

1976-01-01

Upon the occasion of loss of coolant in a nuclear reactor as when a coolant supply or return line breaks, or both lines break, borated liquid coolant from an emergency source is supplied in an amount to absorb heat being generated in the reactor even after the control rods have been inserted. The liquid coolant flows from pressurized storage vessels outside the reactor to an internal manifold from which it is distributed to unused control rod guide thimbles in the reactor fuel assemblies. Since the guide thimbles are mounted at predetermined positions relative to heat generating fuel elements in the fuel assemblies, holes bored at selected locations in the guide thimble walls, sprays the coolant against the reactor fuel elements which continue to dissipate heat but at a reduced level. The cooling water evaporates upon contacting the fuel rods thereby removing the maximum amount of heat (970 BTU per pound of water) and after heat absorption will leave the reactor in the form of steam through the break which is the cause of the accident to help assure immediate core cooldown

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Radiochemical analysis of the first plateout probe from the Fort St. Vrain high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Burnette, R.D.

1982-06-01

This report presents the analysis of radioactive elements on the first plateout probe from the Fort St. Vrain high-temperature gas-cooled reactor. The plateout probe is a device which samples the primary coolant for condensible fission products. Circuit inventories of individual radionuclides are estimated from the probe analysis. The analysis shows that the radioactive contamination in the primary circuit is remarkable low, with activation product concentrations much greater than that of fission products. The analysis demonstrates that the concentrations of the key fission products I-131 and Sr-90 are far below the limits allowed by the technical specification

Power reactors and sub-critical blanket systems with lead and lead-bismuth as coolant and/or target material. Utilization and transmutation of actinides and long lived fission products

International Nuclear Information System (INIS)

2003-05-01

High level radioactive waste disposal is an issue of great importance in the discussion of the sustainability of nuclear power generation. The main contributors to the high radioactivity are the fission products and the minor actinides. The long lived fission products and minor actinides set severe demands on the arrangements for safe waste disposal. Fast reactors and accelerator driven systems (ADS) are under development in Member States to reduce the long term hazard of spent fuel and radioactive waste, taking advantage of their incineration and transmutation capability. Important R and D programmes are being undertaken in many Member States to substantiate this option and advance the basic knowledge in this innovative area of nuclear energy development. The conceptual design of the lead cooled fast reactor concept BREST-OD-300, as well as various other conceptual designs of lead/lead-bismuth cooled fast reactors have been developed to meet enhanced safety and non-proliferation requirements, aiming at both energy production and transmutation of nuclear waste. Some R and D studies indicate that the use of lead and lead-bismuth coolant has some advantages in comparison with existing sodium cooled fast reactor systems, e.g.: simplified design of fast reactor core and BOP, enhanced inherent safety, and easier radwaste management in related fuel cycles. Moreover, various ADS conceptual designs with lead and lead-bismuth as target material and coolant also have been pursued. The results to date are encouraging, indicating that the ADS has the potential to offer an option for meeting the challenges of the back end fuel cycle. During the last decade, there have been substantial advances in several countries with their own R and D programme in the fields of lead/lead-bismuth cooled critical and sub-critical concepts. coolant technology, and experimental validation. In this context, international exchange of information and experience, as well as international

Symbiosis of near breeder HTR's with hybrid fusion reactors

International Nuclear Information System (INIS)

Seifritz, W.

1978-07-01

In this contribution to INFCE a symbiotic fusion/fission reactor system, consisting of a hybrid beam-driven micro-explosion fusion reactor (HMER) and associated high-temperature gas-cooled reactors (HTR) with a coupled fuel cycle, is proposed. This system is similar to the well known Fast Breeder/Near Breeder HTR symbiosis except that the fast fission breeder - running on the U/Pu-cycle in the core and the axial blankets and breeding the surplus fissile material as U-233 in its radial thorium metal or thorium oxide blankets - is replaced by a hybrid micro-explosion DT fusion reactor

Application of Hastelloy X in gas-cooled reactor systems

International Nuclear Information System (INIS)

Brinkman, C.R.; Rittenhouse, P.L.; Corwin, W.R.; Strizak, J.P.; Lystrup, A.; DiStefano, J.R.

1976-10-01

Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data are reported. Properties of concern include tensile, creep, creep-rupture, fatigue, creep-fatigue interaction, subcritical crack growth, thermal stability, and the influence of helium environments with controlled amounts of impurities on these properties. In order to develop these properties in helium environments that are expected to be prototypic of HTGR operating conditions, it was necessary to construct special environmental test systems. Details of construction and operating parameters are described. Interim results from tests designed to determine the above properties are presented. To date a fairly extensive amount of information has been generated on this material at Oak Ridge National Laboratory and elsewhere concerning behavior in air, which is reviewed. However, only limited data are available from tests conducted in helium. Comparisons of the fatigue and subcritical growth behavior in air between Hastelloy X and a number of other structural alloys are given

Core cooling systems

International Nuclear Information System (INIS)

Hoeppner, G.

1980-01-01

The reactor cooling system transports the heat liberated in the reactor core to the component - heat exchanger, steam generator or turbine - where the energy is removed. This basic task can be performed with a variety of coolants circulating in appropriately designed cooling systems. The choice of any one system is governed by principles of economics and natural policies, the design is determined by the laws of nuclear physics, thermal-hydraulics and by the requirement of reliability and public safety. PWR- and BWR- reactors today generate the bulk of nuclear energy. Their primary cooling systems are discussed under the following aspects: 1. General design, nuclear physics constraints, energy transfer, hydraulics, thermodynamics. 2. Design and performance under conditions of steady state and mild transients; control systems. 3. Design and performance under conditions of severe transients and loss of coolant accidents; safety systems. (orig./RW)

Effect of dc-power-system reliability on reactor-shutdown cooling

International Nuclear Information System (INIS)

Kolaczkowski, A.M.; Baranowsky, P.W.; Hickman, J.W.

1981-01-01

The DC power systems in a nuclear power plant provide control and motive power to valves, instrumentation, emergency diesel generators, and many other components and systems during all phases of plant operation including abnormal shutdowns and accident situations. A specific area of concern is the adequacy of the minimum design requirements for DC power systems, particularly with regard to multiple and common cause failures. This concern relates to the application of the single failure criterion for assuring a reliable DC power supply which may be required for the functionability of shutdown cooling systems. The results are presented of a reliability based study performed to assess the adequacy of DC power supply design requirements for currently operating light water reactors with particular attention to shutdown cooling requirements

Simulation of a gas cooled reactor with the system code CATHARE

International Nuclear Information System (INIS)

Bentivoglio, Fabrice; Ruby, Alain; Geffraye, Genevieve; Messie, Anne; Saez, Manuel; Tauveron, Nicolas; Widlund, Ola

2006-01-01

In recent years the CEA has commissioned a wide range of feasibility studies of future advanced nuclear reactors, in particular gas-cooled reactors (GCR). This paper presents an overview of the use of the thermohydraulics code CATHARE in these activities. Extensively validated and qualified for pressurized water reactors, CATHARE has been adapted to deal also with gas-cooled reactor applications. Rather than branching off a separate GCR version of CATHARE, new features have been integrated as independent options in the standard version of the code, respecting the same stringent procedures for documentation and maintenance. CATHARE has evolved into an efficient tool for GCR applications, with first results in good agreement with existing experimental data and other codes. The paper give an example among the studies already carried out with CATHARE with the case of the Very High Temperature Reactor (VHTR) concepts. Current and future activities for experimental validation of CATHARE for GCR applications are also discussed. Short-term validation activities are also included with the assessment of the German utility Oberhausen II. For the long term, CEA has initiated an ambitious experimental program ranging from small scale loops for physical correlations to component technology and system demonstration loops. (authors)

Measurements of actinide-fission product yields in Caliban and Prospero metallic core reactor fission neutron fields

Energy Technology Data Exchange (ETDEWEB)

Casoli, P.; Authier, N. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Laurec, J.; Bauge, E.; Granier, T. [CEA, Centre DIF, 91297 Arpajon (France)

2011-07-01

In the 1970's and early 1980's, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on {sup 233}U, {sup 235}U, and {sup 239}Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energy causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments. (authors)

Water cooled reactor technology: Safety research abstracts no. 1

International Nuclear Information System (INIS)

1990-01-01

The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

Lead-cooled flexible conversion ratio fast reactor

International Nuclear Information System (INIS)

Nikiforova, Anna; Hejzlar, Pavel; Todreas, Neil E.

2009-01-01

Lead-cooled reactor systems capable of accepting either zero or unity conversion ratio cores depending on the need to burn actinides or operate in a sustained cycle are presented. This flexible conversion ratio reactor is a pool-type 2400 MWt reactor coupled to four 600 MWt supercritical CO 2 (S-CO 2 ) power conversion system (PCS) trains through intermediate heat exchangers. The cores which achieve a power density of 112 kW/l adopt transuranic metallic fuel and reactivity feedbacks to achieve inherent shutdown in anticipated transients without scram, and lead coolant in a pool vessel arrangement. Decay heat removal is accomplished using a reactor vessel auxiliary cooling system (RVACS) complemented by a passive secondary auxiliary cooling system (PSACS). The transient simulation of station blackout (SBO) using the RELAP5-3D/ATHENA code shows that inherent shutdown without scram can be accommodated within the cladding temperature limit by the enhanced RVACS and a minimum (two) number of PSACS trains. The design of the passive safety systems also prevents coolant freezing in case all four of the PSACS trains are in operation. Both cores are also shown able to accommodate unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP) accidents using the S-CO 2 PCS.

Systems analysis of a 100-MWe modular liquid metal cooled reactor

International Nuclear Information System (INIS)

Morris, E.E.; Rhow, S.K.; Switick, D.M.

1985-01-01

The response of a 100-MWe modular liquid metal cooled reactor to unprotected loss of flow and/or loss of primary heat removal accidents is analyzed using the systems analysis code SASSYS. The reactor response is tracked for the first 1000 s following a postulated upset in the primary heat removal system. The calculations do not take credit for the functioning of any decay heat removal other than through the secondary system. In addition to the power rating, other features of the reactor are an average sodium temperature rise of 148 K, a sodium void worth (counting the core and upper axial blanket) of 1.89 $, and 3.6 $ of Doppler feedback due to a uniform e-fold fuel temperature increase

Technical evaluation of corium cooling at the reactor cavity

International Nuclear Information System (INIS)

Yang, Soo Hyung; Chan, Eun Sun; Lee, Jae Hun; Lee, Jong In

1998-01-01

To terminate the progression of the severe accident and mitigate the accident consequences, corium cooling has been suggested as one of most important design features considered in the severe accident mitigation. Till now, some kinds of cooling methodologies have been identified and, specially, the corium cooling at the reactor cavity has been considered as one of the most promising cooling methodologies. Moreover, several design requirements related to the corium cooling at the reactor cavity have been also suggested and applied to the design of the next generation reactor. In this study, technical descriptions are briefly described for the important issues related to the corium cooling at the reactor cavity, i.e. cavity area, cavity flooding system, etc., and simple evaluations for those items have been performed considering present technical levels including the experiment and analytical works

Retrofitting the instrumentation and control system of primary cooling circuit from TRIGA INR 14 MW reactor

International Nuclear Information System (INIS)

Preda, M.; Ciocanescu, M.; Ana, E. M.; Cristea, D.

2008-01-01

Activities of retrofitting the instrumentation and control system from TRIGA INR primary cooling circuit consists in replacement of actual system for: - parameter measurement; - safety; - reactor external scramming; - protection, command and supply for electrical elements of the system. This retrofitting project is designed to ensure the necessary features of reactor external safety and for technological parameter measurement. The new safety system of main cooling circuit is completely separated from its operating system and is arranged in a panel assembly in reactor control room. The operating system has the following features: - data acquisition; - parameter value and state of command elements displaying; - command elements on hierarchical levels; - operator information through visual and acoustic alarm. (authors)

CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

International Nuclear Information System (INIS)

Hassan, Yassin; Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

2014-01-01

The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

Energy Technology Data Exchange (ETDEWEB)

Hassan, Yassin [Univ. of Wisconsin, Madison, WI (United Texas A & M Univ., College Station, TX (United States); Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

2014-07-14

The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

Materials compatibility considerations for a fusion-fission hybrid reactor design

International Nuclear Information System (INIS)

DeVan, J.H.; Tortorelli, P.F.

1983-01-01

The Tandem Mirror Hybrid Reactor is a fusion reactor concept that incorporates a fission-suppressed breeding blanket for the production of 233 U to be used in conventional fission power reactors. The present paper reports on compatibility considerations related to the blanket design. These considerations include solid-solid interactions and liquid metal corrosion. Potential problems are discussed relative to the reference blanket operating temperature (490 0 C) and the recycling time of breeding materials (<1 year)

Natural fission reactors - the Oklo phenomenon

International Nuclear Information System (INIS)

Anon.

1989-01-01

Overview describes the discovery of the site, location of the reactors and site geology and discusses the permanence of fission products, nuclear reaction control mechanisms and trace concentrations of elements that act as poisons. (Author)

Fission product behavior in the Molten Salt Reactor Experiment

International Nuclear Information System (INIS)

Compere, E.L.; Kirslis, S.S.; Bohlmann, E.G.; Blankenship, F.F.; Grimes, W.R.

1975-10-01

Essentially all the fission product data for numerous and varied samples taken during operation of the Molten Salt Reactor Experiment or as part of the examination of specimens removed after particular phases of operation are reported, together with the appropriate inventory or other basis of comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct chemical groups. Evidence for fission product behavior during operation over a period of 26 months with 235 U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using 233 U fuel over a period of about 15 months (more than 5100 effective full-power hours)

Lanthanide fission product separation from the transuranics in the integral fast reactor fuel cycle demonstration

International Nuclear Information System (INIS)

Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Ackerman, J.P.

1993-01-01

The Integral Fast Reactor (IFR) is an innovative reactor concept being developed by Argonne National Laboratory. This reactor uses liquid-metal cooling and metallic fuel. Its spent fuel will be reprocessed using a pyrochemical method employing molten salts and liquid metals in an electrofining operation. The lanthanide fission products are a concern during reprocessing because of heating and fuel performance issues, so they must be removed periodically from the system to lessen their impact. The actinides must first be removed form the system before the lanthanides are removed as a waste stream. This operation requires a relatively good lanthanide-actinide separation to minimize both the amount of transuranic material lost in the waste stream and the amount of lanthanides collected when the actinides are first removed. A computer code, PYRO, that models these operations using thermodynamic and empirical data was developed at Argonne and has been used to model the removal of the lanthanides from the electrorefiner after a normal operating campaign. Data from this model are presented. The results demonstrate that greater that 75% of the lanthanides can be separated from the actinides at the end of the first fuel reprocessing campaign using only the electrorefiner vessel

Cooling Performance of Natural Circulation for a Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

On the safety of conceptual fusion-fission hybrid reactors

International Nuclear Information System (INIS)

Kastenberg, W.E.; Okrent, D.; Badham, V.; Caspi, S.; Chan, C.K.; Ferrell, W.J.; Frederking, T.H.K.; Grzesik, J.; Lee, J.Y.; McKone, T.E.; Pomraning, G.C.; Ullman, A.Z.; Ting, T.D.; Kim, Y.I.

1979-01-01

A preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors is presented in this paper. The study and subsequent analysis was largely based upon one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The major potential hazards were found to be: (a) fission products, (b) actinide elements, (c) induced radioactivity, and (d) tritium. As a result of these studies, it appears that highly reliable and even redundent decay heat removal must be provided. Loss of the ability to remove decay heat results in melting of fuel, with ultimate release of fission products and actinides to the containment. In addition, the studies indicate that blankets can be designed which will remain subcritical under extensive changes in both composition and geometry. Magnet safety and the effects of magnetic fields on thermal parameters were also considered. (Auth.)

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

International Nuclear Information System (INIS)

Inaba, Yoshitomo

2008-03-01

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

Fission product behaviour in the primary circuit of an HTR

International Nuclear Information System (INIS)

Decken, C.B. von der; Iniotakis, N.

1981-01-01

The knowledge of fission product behaviour in the primary circuit of a High Temperature Reactor (HTR) is an essential requirement for the estimations of the availability of the reactor plant in normal operation, of the hazards to personnel during inspection and repair and of the potential danger to the environment from severe accidents. On the basis of the theoretical and experimental results obtained at the ''Institute for Reactor Components'' of the KFA Juelich /1/,/2/ the transport- and deposition behaviour of the fission- and activation products in the primary circuit of the PNP-500 reference plant has been investigated thoroughly. Special work had been done to quantify the uncertainties of the investigations and to calculate or estimate the dose rate level at different components of the primary cooling circuit. The contamination and the dose rate level in the inspection gap in the reactor pressure vessel is discussed in detail. For these investigations in particular the surface structure and the composition of the material, the chemical state of the fission products in the cooling gas, the composition of the cooling gas and the influence of dust on the transport- and deposition behaviour of the fission products have been taken into account. The investigations have been limited to the nuclides Ag-110m; Cs-134 and Cs-137

A Review of Previous Research in Direct Energy Conversion Fission Reactors

International Nuclear Information System (INIS)

DUONG, HENRY; POLANSKY, GARY F.; SANDERS, THOMAS L.; SIEGEL, MALCOLM D.

1999-01-01

From the earliest days of power reactor development, direct energy conversion was an obvious choice to produce high efficiency electric power generation. Directly capturing the energy of the fission fragments produced during nuclear fission avoids the intermediate conversion to thermal energy and the efficiency limitations of classical thermodynamics. Efficiencies of more than 80% are possible, independent of operational temperature. Direct energy conversion fission reactors would possess a number of unique characteristics that would make them very attractive for commercial power generation. These reactors would be modular in design with integral power conversion and operate at low pressures and temperatures. They would operate at high efficiency and produce power well suited for long distance transmission. They would feature large safety margins and passively safe design. Ideally suited to production by advanced manufacturing techniques, direct energy conversion fission reactors could be produced more economically than conventional reactor designs. The history of direct energy conversion can be considered as dating back to 1913 when Moseleyl demonstrated that charged particle emission could be used to buildup a voltage. Soon after the successful operation of a nuclear reactor, E.P. Wigner suggested the use of fission fragments for direct energy conversion. Over a decade after Wigner's suggestion, the first theoretical treatment of the conversion of fission fragment kinetic energy into electrical potential appeared in the literature. Over the ten years that followed, a number of researchers investigated various aspects of fission fragment direct energy conversion. Experiments were performed that validated the basic physics of the concept, but a variety of technical challenges limited the efficiencies that were achieved. Most research in direct energy conversion ceased in the US by the late 1960s. Sporadic interest in the concept appears in the literature until this

Mirror reactor studies

International Nuclear Information System (INIS)

Moir, R.W.; Barr, W.L.; Bender, D.J.

1977-01-01

Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

A Two-Phase Cooling Loop for Fission Surface Power Waste Heat Transport, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — Current lunar-based Fission Surface Power (FSP) Systems that will support sustained surface outposts consist of a nuclear reactor with power converters, whose waste...

Development and verification of the LIFE-GCFR computer code for predicting gas-cooled fast-reactor fuel-rod performance

International Nuclear Information System (INIS)

Hsieh, T.C.; Billone, M.C.; Rest, J.

1982-03-01

The fuel-pin modeling code LIFE-GCFR has been developed to predict the thermal, mechanical, and fission-gas behavior of a Gas-Cooled Fast Reactor (GCFR) fuel rod under normal operating conditions. It consists of three major components: thermal, mechanical, and fission-gas analysis. The thermal analysis includes calculations of coolant, cladding, and fuel temperature; fuel densification; pore migration; fuel grain growth; and plenum pressure. Fuel mechanical analysis includes thermal expansion, elasticity, creep, fission-product swelling, hot pressing, cracking, and crack healing of fuel; and thermal expansion, elasticity, creep, and irradiation-induced swelling of cladding. Fission-gas analysis simultaneously treats all major mechanisms thought to influence fission-gas behavior, which include bubble nucleation, resolution, diffusion, migration, and coalescence; temperature and temperature gradients; and fission-gas interaction with structural defects

HLW disposal by fission reactors; calculation of trans-mutation rate and recycle

International Nuclear Information System (INIS)

Mulyanto

1997-01-01

Transmutation of MA (Minor actinide) and LLFPS (long-lived fission products) into stable nuclide or short-lived isotopes by fission reactors seem to become an alternative technology for HLW disposal. in this study, transmutation rate and recycle calculation were developed in order to evaluate transmutation characteristics of MA and LLFPs in the fission reactors. inventory of MA and LLFPs in the transmutation reactors were determined by solving of criticality equation with 1-D cylindrical geometry of multigroup diffusion equations at the beginning of cycle (BOC). transmutation rate and burn-up was determined by solving of depletion equation. inventory of MA and LLFPs was calculated for 40 years recycle. From this study, it was concluded that characteristics of MA and LLFPs in the transmutation reactors can be evaluated by recycle calculation. by calculation of transmutation rate, performance of fission reactor for transmutation of MA or LLFPs can be discussed

Fusion--fission hybrid reactors based on the laser solenoid

International Nuclear Information System (INIS)

Steinhauer, L.C.; Taussig, R.T.; Quimby, D.C.

1976-01-01

Fusion-fission reactors, based on the laser solenoid concept, can be much smaller in scale than their pure fusion counterparts, with moderate first-wall loading and rapid breeding capabilities (1 to 3 tonnes/yr), and can be designed successfully on the basis of classical plasma transport properties and free-streaming end-loss. Preliminary design information is presented for such systems, including the first wall, pulse coil, blanket, superconductors, laser optics, and power supplies, accounting for the desired reactor performance and other physics and engineering constraints. Self-consistent point designs for first and second generation reactors are discussed which illustrate the reactor size, performance, component parameters, and the level of technological development required

Survey on the fusion/fission-hybrid-reactors, a literature review

International Nuclear Information System (INIS)

A survey, based on existing literature, of the work being pursued worldwide on fusion - fission (hybrid) reactor systems is presented. Six areas are reviewed: Plasma physics parameters; Blankets concepts; Fuel cycles; Reactor conceptual designs; Safety and environmental problems; System studies and economic perspectives. Attention has been restricted to systems using magnetically confined plasmas, mainly to mirror and Tokamak - type concepts. The aim is to provide sufficient information, even if not exhaustive, on hybrid reactor concepts in order to help understand what may be expected from their possible development and the ways in which hybrids could affect the future energy scenario. Some concluding remarks are made which represent the personal view of the authors only

Proceedings of the Second Fusion-Fission Energy Systems Review Meeting

Energy Technology Data Exchange (ETDEWEB)

None

1977-11-02

The agenda of the meeting was developed to address, in turn, the following major areas: specific problem areas in nuclear energy systems for application of fusion-fission concepts; current and proposed fusion-fission programs in response to the identified problem areas; target costs and projected benefits associated with fusion-fission energy systems; and technical problems associated with the development of fusion-fission concepts. The greatest emphasis was placed on the characteristics of and problems, associated with fuel producing fusion-fission hybrid reactors.

Passive cooling of a fixed bed nuclear reactor

International Nuclear Information System (INIS)

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

2005-01-01

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

Research activities on high-temperature gas-cooled reactors (HTRs) in the 5. EURATOM RTD Framework programme

International Nuclear Information System (INIS)

Martin-Bermejo, J.; Hugon, M.; Van Goethem, G.

2002-01-01

One of the areas of research of the 'nuclear fission' key action of the 5. EURATOM RTD Framework Programme (FP5) is the safety and efficiency of future systems. The main objective of this area is to investigate and evaluate new or revisited concepts (both reactors and alternative fuels) for nuclear energy that offer potential longer term benefits in terms of cost, safety, waste management, use of fissile material, less risk of diversion and sustainability. Several projects related to high-temperature gas-cooled reactors (HTRs) were retained by the European Commission (EC) services. They address important issues such as HTR fuel technology, HTR fuel cycle, HTR materials, power conversion systems and licensing. Most of these projects have already started and are progressing according to the schedule. They are the initial core of activities of a European Network on 'High-temperature Reactor Technology' (HTR-TN) recently set up by 18 EU organisations. (authors)

Reactor container cooling device

Energy Technology Data Exchange (ETDEWEB)

Ando, Koji; Kinoshita, Shoichiro

1995-11-10

The device of the present invention efficiently lowers pressure and temperature in a reactor container upon occurrence of a severe accident in a BWR-type reactor and can cool the inside of the container for a long period of time. That is, (1) pipelines on the side of an exhaustion tower of a filter portion in a filter bent device of the reactor container are in communication with pipelines on the side of a steam inlet of a static container cooling device by way of horizontal pipelines, (2) a back flow check valve is disposed to horizontal pipelines, (3) a steam discharge valve for a pressure vessel is disposed closer to the reactor container than the joint portion between the pipelines on the side of the steam inlet and the horizontal pipelines. Upon occurrence of a severe accident, when the pressure vessel should be ruptured and steams containing aerosol in the reactor core should be filled in the reactor container, the inlet valve of the static container cooling device is closed. Steams are flown into the filter bent device of the reactor container, where the aerosols can be removed. (I.S.).

Advanced nuclear fuel production by using fission-fusion hybrid reactor

International Nuclear Information System (INIS)

Al-Kusayer, T.A.; Sahin, S.; Abdulraoof, M.

1993-01-01

Efforts are made at the College of Engineering, King Saud University, Riyadh to lay out the main structure of a prototype experimental fusion and fusion-fission (hybrid) reactor blanket in cylindrical geometry. The geometry is consistent with most of the current fusion and hybrid reactor design concepts in respect of the neutronic considerations. Characteristics of the fusion chamber, fusion neutrons and the blanket are provided. The studies have further shown that 1 GWe fission-fusion reactor can produce up to 957 kg/year which is enough to fuel five light water reactors of comparable power. Fuel production can be increased further. 29 refs

Study of risk reduction by improving operation of reactor core isolation cooling system

International Nuclear Information System (INIS)

Watanabe, Yamato; Tazai, Ayuko; Yamagishi, Shohei; Muramatsu, Ken; Muta, Hitoshi

2014-01-01

The Fukushima Daiichi nuclear power plant fell into a station blackout (SBO) due to the earthquake and tsunami in which most of the core cooling systems were disabled. In the units 2 and 3, water injection to the core was performed only by water injection system with turbine driven pumps. In particular, it is inferred from observed plant parameters that the reactor core isolation cooling system (RCIC) continued its operation much longer than it was originally expected (8 hours). Since the preparation of safety measures did not work, the reactor core damaged. With a view to reduce risk of station blackout events in a BWR by accident management, this study investigated the efficacy of operation procedures that takes advantage of RCIC which can be operated with only equipment inside reactor building and does not require an AC power source. The efficacy was assessed in this study by two steps. The first step is a thermal hydraulic analysis with the RETRAN3D code to estimate the potential extension of duration of core cooling by RCIC and the second step is the estimation of time required for recovery of off-site power from experiences at nuclear power stations under the 3.11 earthquake. This study showed that it is possible to implement more reliable measures for accident termination and to greatly reduce the risk of SBO by the installation of accident management measures with use of RCIC for extension of core cooling under SBO conditions. (author)

Analysis for RSG-GAS operational characteristics of reactor cooling system

International Nuclear Information System (INIS)

Nurhappy, T.

1998-01-01

Analysis of operational characteristics of reactor cooling systems (JE01 and PA) is aimed at determining the effects of operation and maintenance patterns to the operational characteristic of the system. Analysis is carried out by virtue of the operating and maintenance data from 1987 to 1997, comprising the operating hours (duration) and data on operating failures of the systems. Results of study show that, either separately or jointly, the operating and maintenance patterns will qualitatively affect the operational characteristic of the systems

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

Science.gov (United States)

Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

1993-01-01

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

Estimation, comparison, and evaluation of advanced fission power reactor generation costs

International Nuclear Information System (INIS)

Waddell, J.D.

1977-01-01

The study compares the high-temperature gas-cooled reactor (HTGR), the gas-cooled fast reactor (GCFR), the molten-salt breeder reactor (MSBR), the light water breeder reactor (LWBR), and the heavy water reactor (HWR) with proposed light water reactors (LWR) and liquid-metal fast breeder reactors (LMFBR). The relative electrical generation costs, including the effects of the introduction of advanced reactor fuel cycles into the U.S. nuclear power economy, were projected through the year 2030. The study utilized the NEEDS computer code which is a simulation of the U.S. nuclear power economy. The future potential electrical generation costs and cumulative consumption of uranium ore were developed using characterizations of the advanced systems. The reactor-fuel cycle characterizations were developed from literature reviews and personal discussions with the proponents of the various systems. The study developed a ranking of the concepts based on generation costs and uranium consumption

Refueling system for the gas-cooled fast breeder reactor

International Nuclear Information System (INIS)

Hawke, B.C.

1980-05-01

Criteria specifically related to the handling of Gas-Cooled Fast Breeder Reactor (GCFR) fuel are briefly reviewed, and the most significant requirements with which the refueling system must comply are discussed. Each component of the refueling system is identified, and a functional description of the fuel handling machine is presented. An illustrated operating sequence describing the various functions involved in a typical refueling cycle is presented. The design status of components and subsystems selected for conceptual development is reviewed, and anticipated refueling time frames are given

Sodium-cooled nuclear reactor

International Nuclear Information System (INIS)

Hammers, H.W.

1982-01-01

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

Design study on sodium-cooled middle-scale modular reactor

International Nuclear Information System (INIS)

Shimakawa, Yoshio; Nibe, Nobuaki; Hori, Toru

2002-05-01

In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled middle-scale modular reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase 2 of the F/S, it is planed to precede a preliminary conceptual design of a sodium-cooled middle-scale modular reactor based on the design of the advanced loop type reactor. Through the design study, it is intended to construct such a plant concept that can show its attraction and competitiveness as a commercialized reactor. This report summarizes the results of the design study on the sodium-cooled middle-scale modular reactor performed in JFY2001, which is the first year of Phase 2. As the construction cost of the sodium-cooled middle-scale modular reactor, which has been constructed in Phase 1, was about 10% higher than that of the sodium-cooled large-scale reactor, a new concept of the middle-scale modular reactor, which is expected to be equal to the large-scale reactor from a viewpoint of economic competitiveness, has been re-constructed based on the design of the advanced loop type reactor. After that, fundamental specifications of main systems and components for the new concept have been set, and critical subjects related to safety, structural integrity, thermal hydraulics, operability, maintainability and economy have been examined and evaluated. As a result of this study, the plant concept of the sodium-cooled middle-scale modular reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000 yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

SSTAR: The US lead-cooled fast reactor (LFR)

International Nuclear Information System (INIS)

Smith, Craig F.; Halsey, William G.; Brown, Neil W.; Sienicki, James J.; Moisseytsev, Anton; Wade, David C.

2008-01-01

It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. It is a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation goals, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the US Generation IV lead-cooled fast reactor system

Proposal of a system for fuel elements inspection of CDTN TRIGA nuclear reactor

International Nuclear Information System (INIS)

Rodrigues, Rogerio Rivail; Mesquita, Amir Zacarias

2013-01-01

The CDTN has in its facilities a TRIGA-type nuclear reactor. The reactor's cooling water must be treated and managed with the goal of keeping its low conductivity to minimize corrosion of the reactor components, mainly of fuel elements (FE), and reduce the level of radioactivity. The aim of this paper is to present a proposal for the development of a system for verification of some possible leaks in FE nuclear research reactors, based on the sipping test. This type of testing is a way to check for leaks of fission products from fuel element of nuclear research reactor. In the future, when the test will do, it will have a correlation between the components found in the reactor cooling water pool and integrity of nuclear fuel elements. The device development and its application will be presented here, covering results that were not previously investigated yet, giving originality to this project. (author)

Emergency core cooling system

International Nuclear Information System (INIS)

Abe, Nobuaki.

1993-01-01

A reactor comprises a static emergency reactor core cooling system having an automatic depressurization system and a gravitationally dropping type water injection system and a container cooling system by an isolation condenser. A depressurization pipeline of the automatic depressurization system connected to a reactor pressure vessel branches in the midway. The branched depressurizing pipelines are extended into an upper dry well and a lower dry well, in which depressurization valves are disposed at the top end portions of the pipelines respectively. If loss-of-coolant accidents should occur, the depressurization valve of the automatic depressurization system is actuated by lowering of water level in the pressure vessel. This causes nitrogen gases in the upper and the lower dry wells to transfer together with discharged steams effectively to a suppression pool passing through a bent tube. Accordingly, the gravitationally dropping type water injection system can be actuated faster. Further, subsequent cooling for the reactor vessel can be ensured sufficiently by the isolation condenser. (I.N.)

Tokamak hybrid thermonuclear reactor for the production of fissionable fuel and electric power

International Nuclear Information System (INIS)

Velikhov, E.P.; Glukhikh, V.A.; Gur'ev, V.V.

1978-01-01

The results of feasibility studies of a tokamak- based hybrid reactor concept are presented. The system selected has a D-T plasma volume of 575 m 3 with additional plasma heating by injection of fast neutral particles. The method of heating makes it possible to achieve an economical two-component tokamak regime at ntau=(4-6)x10 13 sxcm -3 , i e. far below the Lawson criterion. Plasma and vacuum chamber are surrounded by a blanket where fissionable plutonium is produced and heat transformed into electric power is generated. Major plasma-neutron-physical characteristics of the 6905 MWth (2500 MWe) reactor and its electromagnetic system are presented. Evaluations show that the hybrid reactor can produce about 800 kg of Pu per 1GWth/yr as compared to 70-150 kg of Pu for fast breeder reactors. The increased Pu production rate is the major merit of the concept promising for both power generation and fuelling thermal fission reactions

Role of fission-reactor-testing capabilities in the development of fusion technology

International Nuclear Information System (INIS)

Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Schmunk, R.E.; Takata, M.L.; Watts, K.D.

1981-01-01

Testing of fusion materials and components in fission reactors will be increasingly important in the future due to the near-term lack of fusion engineering test devices, and the long-term high demand for testing when fusion reactors become available. Fission testing is capable of filling many gaps in fusion reactor design information, and thus should be aggressively pursued. EG and G Idaho has investigated the application of fission testing in three areas, which are discussed in this paper. First, we investigated radiation damage to magnet insulators. This work is now continuing with the use of an improved test capsule. Second, a study was performed which indicated that a fission-suppressed hybrid blanket module could be effectively tested in a reactor such as the Engineering Test Reactor (ETR), closely reproducing the predicted performance in a fusion environment. Finally, we explored a conceptual design for a fission-based Integrated Test Facility (ITF), which can accommodate entire First Wall/Blanket (FW/B) modules for testing in a nuclear environment, simultaneously satisfying many of the FW/B test requirements. This ITF can provide a cyclic neutron/gamma flux, as well as the necessary module support functions

Euratom innovation in nuclear fission: Community research in reactor systems and fuel cycles

International Nuclear Information System (INIS)

Goethem, G. van; Hugon, M.; Bhatnagar, V.; Manolatos, P.; Deffrennes, M.

2007-01-01

The following questions are naturally at the heart of the current Euratom research and training framework programme:(1)What are the challenges facing the European Union nuclear fission research community in the short (today), medium (2010) and long term (2040)? (2)What kind of research and technological development (RTD) does Euratom offer to respond to these challenges, in particular in the area of reactor systems and fuel cycles? In the general debate about energy supply technologies there are challenges of both a scientific and technological (S/T) as well as an economic and political (E/P) nature. Though the Community research programme acts mainly on the former, there is nevertheless important links with Community policy. These not only exist in the specific area of nuclear policy, but also more generally as is depicted in the following figure. It is shown in the particular area of nuclear fission, to what extent Euratom research, education and innovation ('Knowledge Triangle' in above figure) respond to the following long-term criteria: (1) sustainability, (2) economics, (3) safety, and (4) proliferation resistance. Research and innovation in nuclear fission technology has broad and extended geographical, disciplinary and time horizons:- the community involved extends to all 25 EU Member States and beyond; - the research assembles a large variety of scientific disciplines; - three generations of nuclear power technologies (called II, III and IV) are involved, with the timescales extending from now to around the year 2040. To each of these three generations, a couple of challenges are associated (six in total):- Generation II (1970-2000, today): security of supply+environmental compatibility; - Generation III (around 2010): enhanced safety and competitiveness (economics); - Generation IV (around 2040): cogeneration of heat and power, and full recycling. At the European Commission (EC), the research related to nuclear reactor systems and fuel cycles is

Properties of Fission-Product decay heat from Minor-Actinide fissioning systems

International Nuclear Information System (INIS)

Oyamatsu, Kazuhiro; Mori, Hideki

2000-01-01

The aggregate Fission-Product (FP) decay heat after a pulse fission is examined for Minor Actinide (MA) fissiles 237 Np, 241 Am, 243 Am, 242 Cm and 244 Cm. We find that the MA decay heat is comparable but smaller than that of 235 U except for cooling times at about 10 8 s (approx. = 3 y). At these cooling times, either the β or γ component of the FP decay heat for these MA's is substantially larger than the one for 235 U. This difference is found to originate from the cumulative fission yield of 106 Ru (T 1/2 = 3.2x10 7 s). This nuclide is the parent of 106 Rh (T 1/2 = 29.8 s) which is the dominant source of the decay heat at 10 8 s (approx. = 3 y). The fission yield is nearly an increasing function of the fissile mass number so that the FP decay heat is the largest for 244 Cm among the MA's at the cooling time. (author)

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

International Nuclear Information System (INIS)

Schikorr, W.M.

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

A Neutronic Feasibility Study on the Recycling of an Oxide Fuel in Sodium-Cooled Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Roh, Gyu Hong; Choi, Hang Bok

2006-06-15

Neutronic feasibility was implemented for the recycling of a mixed oxide fuel in sodium-cooled fast reactors (SFR) through a thermal/mechanical dry process, which is recognized as one of the most proliferation- resistant recycling processes. In order to assess the applicability of a simple dry process which is not capable of completely removing all the fission products from a spent fuel, sensitivity calculations were performed for the reactor physics parameters with a dependency on the fission product removal rate of the recycled spent fuel. The equilibrium core calculations were performed by the REBUS-3 code for a BN-600 core without blanket fuels and a modified core with an increased fuel volume fraction. The reactor performance parameters such as the transuranic content, breeding ratio, peak linear power, burnup reactivity swing and reactivity coefficients were calculated for an equilibrium core under a fixed fuel management scheme. The results showed that a recycling of the oxide fuel in the SFR is feasible if the fission products are removed by more than 70% through the dry process as far as the material balance is concerned. However the physics analysis also showed that some of the physics design parameters are slightly deteriorated. The results of this study indicate that the recycling characteristics can be improved if the dry process can remove more fission products, and the reactor configuration is further optimized or the spent fuel composition is adjusted.

A Neutronic Feasibility Study on the Recycling of an Oxide Fuel in Sodium-Cooled Fast Reactors

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok

2006-06-01

Neutronic feasibility was implemented for the recycling of a mixed oxide fuel in sodium-cooled fast reactors (SFR) through a thermal/mechanical dry process, which is recognized as one of the most proliferation- resistant recycling processes. In order to assess the applicability of a simple dry process which is not capable of completely removing all the fission products from a spent fuel, sensitivity calculations were performed for the reactor physics parameters with a dependency on the fission product removal rate of the recycled spent fuel. The equilibrium core calculations were performed by the REBUS-3 code for a BN-600 core without blanket fuels and a modified core with an increased fuel volume fraction. The reactor performance parameters such as the transuranic content, breeding ratio, peak linear power, burnup reactivity swing and reactivity coefficients were calculated for an equilibrium core under a fixed fuel management scheme. The results showed that a recycling of the oxide fuel in the SFR is feasible if the fission products are removed by more than 70% through the dry process as far as the material balance is concerned. However the physics analysis also showed that some of the physics design parameters are slightly deteriorated. The results of this study indicate that the recycling characteristics can be improved if the dry process can remove more fission products, and the reactor configuration is further optimized or the spent fuel composition is adjusted

Preliminary Sensitivity Study on Gas-Cooled Reactor for NHDD System Using MARS-GCR

International Nuclear Information System (INIS)

Lee, Seung Wook; Jeong, Jae Jun; Lee, Won Jae

2005-01-01

A Gas-Cooled Reactor (GCR) is considered as one of the most outstanding tools for a massive hydrogen production without CO 2 emission. Till now, two types of GCR are regarded as a viable nuclear reactor for a hydrogen production: Prismatic Modular Reactor (PMR), Pebble Bed Reactor (PBR). In this paper, a preliminary sensitivity study on two types of GCR is carried out by using MARS-GCR to find out the effect on the peak fuel and reactor pressure vessel (RPV) temperature, with varying the condition of a reactor inlet, outlet temperature, and system pressure for both PMR and PBR

Reactor system

International Nuclear Information System (INIS)

Miyano, Hiroshi; Narabayashi, Naoshi.

1990-01-01

The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

Detection of coolant void in lead-cooled fast reactors

International Nuclear Information System (INIS)

Wolniewicz, Peter; Håkansson, Ane; Jansson, Peter

2015-01-01

Highlights: • We model the ALFRED LFR using different Monte-Carlo codes. • We study the impact on coolant void on the fission cross section in fission chambers. • We develop a methodology to detect coolant void. • We study the impact of detector fissile coating burn-up. • We conclude that the developed methodology may be an attractive complement to LFR monitoring. - Abstract: Previous work (Wolniewicz et al., 2013) has indicated that using fission chambers coated with 242 Pu and 235 U, respectively, can provide the means of detecting changes in the neutron flux that are connected to coolant density changes in a small lead-cooled fast reactor. Such density changes may be due to leakages of gas into the coolant, which, over time, may coalesce to large bubbles implying a high risk of causing severe damage of the core. By using the ratio of the information provided by the two types of detectors a quantity is obtained that is sensitive to these density changes and, to the first order approximation, independent of the power level of the reactor. In this work we continue the investigation of this proposed methodology by applying it to the Advanced LFR European Demonstrator (ALFRED) and using realistic modelling of the neutron detectors. The results show that the methodology may be used to detect density changes indicating the initial stages of a coalescence process that may result in a large bubble. Also, it is shown that under certain circumstances, large bubbles passing through the core could be detected with this methodology

Chemistry of fission product iodine under nuclear reactor accident conditions

International Nuclear Information System (INIS)

Malinauskas, A.P.; Bell, J.T.

1986-01-01

The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs

Unconventional liquid metal cooled fast reactors

International Nuclear Information System (INIS)

Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.

1989-06-01

This report describes the rationale for, design of and analytical studies on an unconventional sodium-cooled power reactor, called the Trench Reactor. It derives its name from the long, narrow sodium pool in which the reactor is placed. Unconventional features include: pool shape; reactor shape (also long and narrow); reflector control; low power density; hot-leg primary pumping; absence of a cold sodium pool; large core boxes rather than a large number of subassemblies; large diameter metal fuel; vessel suspension from cables; and vessel cooling by natural circulation of building atmosphere (nitrogen) at all times. These features all seem feasible. They result in a system that is capable of at least a ten year reload interval and shows good safety through direct physical response to loss-of-heat-sink, loss-of-flow and limited-reactivity nuclear transients. 43 figs., 43 tabs

Assessment of the dry process fuel sodium-cooled fast reactors

Energy Technology Data Exchange (ETDEWEB)

Roh, Gyu Hong; Choi, Hang Bok

2004-04-01

The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed.

Assessment of the dry process fuel sodium-cooled fast reactors

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok

2004-04-01

The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed

Challenges and innovative technologies on fuel handling systems for future sodium-cooled fast reactors

International Nuclear Information System (INIS)

Chassignet, Mathieu; Dumas, Sebastien; Penigot, Christophe; Prele, Gerard; Capitaine, Alain; Rodriguez, Gilles; Sanseigne, Emmanuel; Beauchamp, Francois

2011-01-01

The reactor refuelling system provides the means of transporting, storing, and handling reactor core subassemblies. The system consists of the facilities and equipment needed to accomplish the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor vessel (primary vessel, storage, and final cooling before going to reprocessing), its construction cost, and its availability factor. Fuel handling design must take into account various items and in particular operating strategies such as core design and management and core configuration. Moreover, the FHS will have to cope with safety assessments: a permanent cooling strategy to prevent fuel clad rupture, plus provisions to handle short-cooled fuel and criteria to ensure safety during handling. In addition, the handling and elimination of residual sodium must be investigated; it implies specific cleaning treatment to prevent chemical risks such as corrosion or excess hydrogen production. The objective of this study is to identify the challenges of a SFR fuel handling system. It will then present the range of technical options incorporating innovative technologies under development to answer the GENERATION IV SFR requirements. (author)

Transmutation of Tc-99 in fission reactors

International Nuclear Information System (INIS)

Kloosterman, J.L.; Li, J.M.

1994-12-01

Transmutation of Tc-99 in three different types of fission reactors is considered: A heavy water reactor, a fast reactor and a light water reactor. For the first type a CANDU reactor was chosen, for the second one the Superphenix reactor, and for the third one a PWR. The three most promising Tc-99 transmuters are the fast reactor with a moderated subassembly in the inner core, a fast reactor with a non-moderated subassembly in the inner core, and a heavy water reactor with Tc-99 target pins in the moderator between the fuel bundles. Transmutation half lives of 15 to 25 years can be achieved, with yearly transmuted Tc-99 masses of about 100 kg at a thermal reactor power of about 3000 MW. (orig.)

Integral measurement of fission products capture in fast breeder reactors

International Nuclear Information System (INIS)

Martin Deidier, Loick.

1979-12-01

For the SUPERPHENIX reactor project, it was necessary to know fission products capture with about 10% accuracy in the fast breeder reactor spectra. In this purpose, integral measurements have been carried out on the main separated products by different experimental technics (oscillation, activation and irradiation methods), but particularly on irradiated fuel pins from RAPSODIE and PHENIX reactors in order to directly obtain total effect of fission products. Same tendencies have been observed for both enriched uranium fuel and LMFBR characteristic plutonium fuel. All experimental results have been introduced in CARNAVAL cross section set [fr

Influence of remaining fission products in low-decontaminated fuel on reactor core characteristics

International Nuclear Information System (INIS)

Ohki, Shigeo

2002-07-01

Design study of core, fuel and related fuel cycle system with low-decontaminated fuel has been performed in the framework of the feasibility study (F/S) on commercialized fast reactor cycle systems. This report summarizes the influence on core characteristics of remaining fission products (FPs) in low-decontaminated fuel related to the reprocessing systems nominated in F/S phase I. For simple treatment of the remaining FPs in core neutronics calculation the representative nuclide method parameterized by the FP equivalent coefficient and the FP volume fraction was developed, which enabled an efficient evaluation procedure. As a result of the investigation on the sodium cooled fast reactor with MOX fuel designed in fiscal year 1999, it was found that the pyrochemical reprocessing with molten salt (the RIAR method) brought the largest influence. Nevertheless, it was still within the allowable range. Assuming an infinite-times recycling, the alternations in core characteristics were evaluated as follows: increment of burnup reactivity by 0.5%Δk/kk', decrement of breeding ratio by 0.04, increment of sodium void reactivity by 0.1x10 -2 Δk/kk' and decrement of Doppler constant (in absolute value) by 0.7x10 -3 Tdk/dT. (author)

Heat transport and afterheat removal for gas cooled reactors under accident conditions

International Nuclear Information System (INIS)

2001-01-01

The Co-ordinated Research Project (CRP) on Heat Transport and Afterheat Removal for Gas Cooled Reactors Under Accident Conditions was organized within the framework of the International Working Group on Gas Cooled Reactors (IWGGCR). This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs) and supports the conduct of these activities. Advanced GCR designs currently being developed are predicted to achieve a high degree of safety through reliance on inherent safety features. Such design features should permit the technical demonstration of exceptional public protection with significantly reduced emergency planning requirements. For advanced GCRs, this predicted high degree of safety largely derives from the ability of the ceramic coated fuel particles to retain the fission products under normal and accident conditions, the safe neutron physics behaviour of the core, the chemical stability of the core and the ability of the design to dissipate decay heat by natural heat transport mechanisms without reaching excessive temperatures. Prior to licensing and commercial deployment of advanced GCRs, these features must first be demonstrated under experimental conditions representing realistic reactor conditions, and the methods used to predict the performance of the fuel and reactor must be validated against these experimental data. Within this CRP, the participants addressed the inherent mechanisms for removal of decay heat from GCRs under accident conditions. The objective of this CRP was to establish sufficient experimental data at realistic conditions and validated analytical tools to confirm the predicted safe thermal response of advance gas cooled reactors during accidents. The scope includes experimental and analytical investigations of heat transport by natural convection conduction and thermal

High Temperature Gas Cooled Reactor Fuels and Materials

International Nuclear Information System (INIS)

2010-03-01

At the third annual meeting of the technical working group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), held in Vienna, in 2004, it was suggested 'to develop manuals/handbooks and best practice documents for use in training and education in coated particle fuel technology' in the IAEA's Programme for the year 2006-2007. In the context of supporting interested Member States, the activity to develop a handbook for use in the 'education and training' of a new generation of scientists and engineers on coated particle fuel technology was undertaken. To make aware of the role of nuclear science education and training in all Member States to enhance their capacity to develop innovative technologies for sustainable nuclear energy is of paramount importance to the IAEA Significant efforts are underway in several Member States to develop high temperature gas cooled reactors (HTGR) based on either pebble bed or prismatic designs. All these reactors are primarily fuelled by TRISO (tri iso-structural) coated particles. The aim however is to build future nuclear fuel cycles in concert with the aim of the Generation IV International Forum and includes nuclear reactor applications for process heat, hydrogen production and electricity generation. Moreover, developmental work is ongoing and focuses on the burning of weapon-grade plutonium including civil plutonium and other transuranic elements using the 'deep-burn concept' or 'inert matrix fuels', especially in HTGR systems in the form of coated particle fuels. The document will serve as the primary resource materials for 'education and training' in the area of advanced fuels forming the building blocks for future development in the interested Member States. This document broadly covers several aspects of coated particle fuel technology, namely: manufacture of coated particles, compacts and elements; design-basis; quality assurance/quality control and characterization techniques; fuel irradiations; fuel

Design study on sodium-cooled large-scale reactor

International Nuclear Information System (INIS)

Shimakawa, Yoshio; Nibe, Nobuaki; Hori, Toru

2002-05-01

In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled large-scale reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase 2 of the F/S, it is planed to precede a preliminary conceptual design of a sodium-cooled large-scale reactor based on the design of the advanced loop type reactor. Through the design study, it is intended to construct such a plant concept that can show its attraction and competitiveness as a commercialized reactor. This report summarizes the results of the design study on the sodium-cooled large-scale reactor performed in JFY2001, which is the first year of Phase 2. In the JFY2001 design study, a plant concept has been constructed based on the design of the advanced loop type reactor, and fundamental specifications of main systems and components have been set. Furthermore, critical subjects related to safety, structural integrity, thermal hydraulics, operability, maintainability and economy have been examined and evaluated. As a result of this study, the plant concept of the sodium-cooled large-scale reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

Radioactive material transport in sodium-cooled nuclear reactors

International Nuclear Information System (INIS)

Brehm, W.F.; McGuire, J.C.; Colburn, R.P.; Maffei, H.P.; Olson, W.H.

1980-03-01

Trapping devices which remove nuclides from the sodium stream in pre-selected locations away from maintenance areas have been developed and proven successful in in-reactor testing. The release of corrosion product radionuclides as a function of system temperature and oxygen content has been quantitatively evaluated. Ongoing work concentrates on further in-reactor testing of radionuclide removal devices, and characterization of fission product release and deposition from fuel pins with breached-cladding

Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

International Nuclear Information System (INIS)

Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

1986-05-01

In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant

Status of and prospects for gas-cooled reactors

International Nuclear Information System (INIS)

1984-01-01

The IAEA International Working Group on Gas-Cooled Reactors (IWGGCR) (see Annex I), which was established in 1978, recommended to the Agency that a report be prepared in order to provide an up-to-date summary of gas-cooled reactor technology. The present Technical Report is based mainly on submissions of Member Countries of the IWGGCR and consists of four main sections. Beside some general information about the gas-cooled reactor line, section 1 contains a description of the incentives for the development and deployment of gas-cooled reactors in various Agency Member States. These include both electricity generation and process steam and process heat production for various branches of industry. The historical development of gas-cooled reactors is reviewed in section 2. In this section information is provided on how, when and why gas-cooled reactors have been developed in various Agency Member States and, in addition, a detailed description of the different gas-cooled reactor lines is presented. Section 3 contains information about the technical status of gas-cooled reactors and their applications. Gas-cooled reactors that are under design or construction or in operation are listed and shortly described, together with an outlook for future reactor designs. In this section the various applications for gas-cooled reactors are described in detail. These include both electricity generation and process steam and process heat production. The last section (section 4) is entitled ''Special features of gas-cooled reactors'' and contains information about the technical performance, fuel utilization, safety characteristics and environmental impact, such as radiation exposure and heat rejection

Research towards ultrasonic systems to assist in-vessel manipulations in liquid metal cooled reactors

International Nuclear Information System (INIS)

Dierckx, Marc; Van-Dyck, Dries

2013-06-01

We describe the state of the art of the research towards ultrasonic measurement methods for use in lead-bismuth cooled liquid metal reactors. Our current research activities are highly focused on specific tasks in the MYRRHA system, which is a fast spectrum research reactor cooled with the eutectic mixture of lead and bismuth (LBE) and is conceived as an accelerator driven system capable of operating in both sub-critical and critical mode. As liquid metal is opaque to light, normal visual feedback during fuel manipulations in the reactor vessel is not available and must therefore be replaced by a system that is not hindered by the opacity of the coolant. In this respect ultrasonic measurement techniques have been proposed and even developed in the past for operation in sodium cooled reactors. To our knowledge, no such systems have ever been deployed in lead based reactors and we are the first to have a research program in this direction as will be detailed in this paper. We give an overview of the acoustic properties of LBE and compare them with the properties of sodium and water to theoretically show the feasibility of ultrasonic systems operating in LBE. In the second part of the paper we discuss the results of the validation experiments in water and LBE. A typical scene is ultrasonically probed by a mechanical scanning system while the signals are processed to render a 3D visualization on a computer screen. It will become clear that mechanical scanning is capable of producing acceptable images but that it is a time consuming process that is not fit to solve the initial task to providing feedback during manipulations in the reactor vessel. That is why we propose to use several dedicated ultrasonic systems each adapted to a specific task and capable to provide real-time feedback of the ongoing manipulations, as is detailed in the third and final part of the paper. (authors)

High temperature helium-cooled fast reactor (HTHFR)

International Nuclear Information System (INIS)

Karam, R.A.; Blaylock, Dwayne; Burgett, Eric; Mostafa Ghiaasiaan, S.; Hertel, Nolan

2006-01-01

Scoping calculations have been performed for a very high temperature (1000 o C) helium-cooled fast reactor involving two distinct options: (1) using graphite foam into which UC (12% enrichment) is embedded into a matrix comprising UC and graphite foam molded into hexagonal building blocks and encapsulated with a SiC shell covering all surfaces, and (2) using UC only (also 12% enrichment) molded into the same shape and size as the foam-UC matrix in option 1. Both options use the same basic hexagonal fuel matrix blocks to form the core and reflector. The reflector contains natural uranium only. Both options use 50 μm SiC as a containment shell for fission product retention within each hexagonal block. The calculations show that the option using foam (option 1) would produce a reactor that can operate continuously for at least 25 years without ever adding or removing any fuel from the reactor. The calculations show further that the UC only option (option 2) can operate continually for 50 years without ever adding or removing fuel from the reactor. Doppler and loss of coolant reactivity coefficients were calculated. The Doppler coefficient is negative and much larger than the loss of coolant coefficient, which was very small and positive. Additional progress on and development of the two concepts are continuing

Maintenance of fission and fusion reactors. 10. workshop on fusion reactor engineering

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-02-01

This report contains copies of OHP presented at the title meeting. The presented topics are as follows, maintenance of nuclear power plants and ITER, exchange of shroud in BWR type reactors, deterioration of fission and fusion reactor materials, standards of pressure vessels, malfunction diagnosis method with neural network. (J.P.N.)

77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

Science.gov (United States)

2012-06-15

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling... for public comment draft regulatory guide (DG), DG-1277, ``Initial Test Program of Emergency Core..., entitled, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors,'' is...

Progress in design, research and development and testing of safety systems for advanced water cooled reactors. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1996-04-01

The meeting covered the following topics: Developments in design of safety-related heat removal components and systems for advanced water cooled reactors; status of test programmes on heat removal components and systems of new designs; range of validity and extrapolation of test results for the qualification of design/licensing computer models and codes for advanced water cooled reactors; future needs and trends in testing of safety systems for advanced water cooled reactors. Tests of heat removal safety systems have been conducted by various groups supporting the design, testing and certification of advanced water cooled reactors. The Technical Committee concluded that the reported test results generally confirm the predicted performance features of the advanced designs. Refs, figs, tabs

Progress in design, research and development and testing of safety systems for advanced water cooled reactors. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-04-01

The meeting covered the following topics: Developments in design of safety-related heat removal components and systems for advanced water cooled reactors; status of test programmes on heat removal components and systems of new designs; range of validity and extrapolation of test results for the qualification of design/licensing computer models and codes for advanced water cooled reactors; future needs and trends in testing of safety systems for advanced water cooled reactors. Tests of heat removal safety systems have been conducted by various groups supporting the design, testing and certification of advanced water cooled reactors. The Technical Committee concluded that the reported test results generally confirm the predicted performance features of the advanced designs. Refs, figs, tabs.

Thermal Energetic Reactor with High Reproduction of Fission Materials

Directory of Open Access Journals (Sweden)

Vladimir M. Kotov

2012-01-01

On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.

High-temperature and breeder reactors - economic nuclear reactors of the future

International Nuclear Information System (INIS)

Djalilzadeh, A.M.

1977-01-01

The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

Brief review of the fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Tenney, F.H.

1977-01-01

Much of the conceptual framework of present day fusion-fission hybrid reactors is found in the original work of the early 1950's. Present day motivations for development are quite different. The role of the hybrid reactor is discussed as well as the current activities in the development program

The study of two, three and four dimensional nonlinear dynamics of nuclear fission reactors and effective parameters on its behaviour

International Nuclear Information System (INIS)

Tajik, M.; Ghasemizad, A.

2008-01-01

In this research, new physical fission reactor parameters which have very sensitive effects on the qualitative behavior of a reactor, are introduced. Therefore, the two, the nonlinear dynamics of two, three and four dimensional, considering almost the effective parameters are formulated for describing nuclear fission reactor systems. Using both analytical and numerical methods, the stability and instability of the given dynamical equations and the conditions of stability are studied in these systems. We have shown that the two parameters of the mean energy residence time in fuel and coolant and also their ratios have the most qualitative effects on the dynamical behaviour of a typical nuclear fission reactor. Increasing or decreasing of these parameters from a captain limit can lead to stability or un stability in a given system

Sealing of leaks in the bioshield cooling system of three research reactors

International Nuclear Information System (INIS)

May, R.; Taylor, M.F.

1995-01-01

Water leaks have occurred in the bioshield cooling system of three research reactors. These leaks have been plugged with a sealant based on a blend of a water-based resin and a bentonite-type clay originally developed for sealing similar leaks on power reactors. The mechanism of sealing and development testing of the sealant are described. Application of the sealant to the three reactors sealed the leaks. However, unlike experience with leaks in steel and aluminium systems, some leaks reappeared after several months service - albeit at a leak rate only a very small fraction of the original leak rate. The recurrent defects were readily retreated with sealant and hence, in these instances, the treatment is an effective maintenance procedure for any ageing reactor rather than a permanent cure. (orig.)

Chemical immobilization of fission products reactive with nuclear reactor components

International Nuclear Information System (INIS)

Grossman, L.N.; Kaznoff, A.I.; Clukey, H.V.

1975-01-01

This invention teaches a method of immobilizing deleterious fission products produced in nuclear fuel materials during nuclear fission chain reactions through the use of additives. The additives are disposed with the nuclear fuel materials in controlled quantities to form new compositions preventing attack of reactor components, especially nuclear fuel cld, by the deleterious fission products. (Patent Office Record)

RCCS Experiments and Validation for High Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Chang Oh; Cliff Davis; Goon C. Park

2007-01-01

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

The continuous fuel cycle model and the gas cooled fast reactor

International Nuclear Information System (INIS)

Christie, Stuart; Lathouwers, Danny; Kloosterman, Jan Leen; Hagen, Tim van der

2011-01-01

The gas cooled fast reactor (GFR) is one of the generation IV designs currently being evaluated for future use. It is intended to behave as an isobreeder, producing the same amount of fuel as it consumes during operation. The actinides in the fuel will be recycled repeatedly in order to minimise the waste output to fission products only. Striking the balance of the fissioning of various actinides against transmutation and decay to achieve these goals is a complex problem. This is compounded by the time required for burn-up modelling, which can be considerable for a single cycle, and even longer for studies of fuel evolution over many cycles. The continuous fuel cycle model approximates the discrete steps of loading, operating and unloading a reactor as continuous processes. This simplifies the calculations involved in simulating the behaviour of the fuel, reducing the time needed to model the changes to the fuel composition over many cycles. This method is used to study the behaviour of GFR fuel over many cycles and compared to results obtained from direct calculations. The effects of varying fuel cycle properties such as feed material, recycling of additional actinides and reprocessing losses are also investigated. (author)

Power Nuclear Reactors: technology and innovation for development in future

International Nuclear Information System (INIS)

Suarez Antola, R.

2009-01-01

The conference is about some historicals task of the fission technology as well as many types of Nuclear Reactors. Enrichment of fuel, wastes, research reactors and power reactors, a brief advertisment about Uruguay electric siystem and power generation, energetic worldwide, proliferation, safety reactors, incidents, accidents, Three-Mile Island accident, Chernobil accident, damages, risks, classification and description of Power reactors steam generation, nuclear reactor cooling systems, future view

Corrosion inhibition measures in primary cooling water system during refurbishment of Cirus, re-commissioning and subsequent operation

International Nuclear Information System (INIS)

Rai, K.K.; Ramesh, N.; Sharma, R.C.

2008-01-01

Cirus is a 40 MWth, heavy water moderated, demineralized light water cooled, natural uranium fuelled research reactor. Reactor was commissioned in year 1960 and operated satisfactorily till 1990. After that availability factor started decreasing mainly due to equipment outage exhibiting signs of ageing. Based upon systematic ageing studies and assessment of condition of systems, structures and components, a refurbishment plan including safety upgrades was drawn up. Reactor was shut down in October 1997 for execution of jobs. After completion of refurbishment jobs reactor was started back in October 2002 and power operation was achieved in 2003. Primary cooling water (PCW) system consists of re-circulating pumps, heat exchangers, expansion tank, piping, valves, emergency storage reservoir (Ball Tank) and other components. Normally the fission heat from fuel is removed by re-circulating coolant in closed loop and transferred to seawater via heat exchangers. In case of outage of pumps, shut down cooling is provided by flow of water from Ball Tank under gravity to the underground dump tanks. The dissolved oxygen is maintained below 2 ppm and pH is maintained neutral to minimize corrosion of fuel cladding (Aluminum). This paper highlights the experience gained during segmentation of primary cooling water pipelines for pressure testing, measures taken to corrosion inhibition of primary cooling water lines to permit execution of refurbishment jobs, inspections and actions taken to repair/replace the corroded PCW pipe line segments, observations regarding corrosion related failures, re-commissioning of the system after refurbishment, assessment for safe reactor operation and experience during power operation. (author)

Determination of the fission coefficients in thermal nuclear reactors for antineutrino detection

Energy Technology Data Exchange (ETDEWEB)

Araujo, Lenilson M. [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Cabral, Ronaldo G., E-mail: rgcabral@ime.eb.b [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil); Anjos, Joao C.C. dos, E-mail: janjos@cbpf.b [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil). Dept. GLN - G

2011-07-01

The nuclear reactors in operation periodically need to change their fuel. It is during this process that these reactors are more vulnerable to occurring of several situations of fuel diversion, thus the monitoring of the nuclear installations is indispensable to avoid events of this nature. Considering this fact, the most promissory technique to be used for the nuclear safeguard for the nonproliferation of nuclear weapons, it is based on the detection and spectroscopy of antineutrino from fissions that occur in the nuclear reactors. The detection and spectroscopy of antineutrino, they both depend on the single contribution for the total number of fission of each actinide in the core reactor, these contributions receive the name of fission coefficients. The goal of this research is to show the computational and mathematical modeling used to determinate these coefficients for PWR reactors. (author)

Structural materials issues for the next generation fission reactors

Science.gov (United States)

Chant, I.; Murty, K. L.

2010-09-01

Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

Energy Technology Data Exchange (ETDEWEB)

Chang Oh

2006-03-01

The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

International Nuclear Information System (INIS)

Chang Oh

2006-01-01

The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 900 C and operational fuel temperatures above 1250 C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR's higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gases (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

Technical Bases to Consider for Performance and Demonstration Testing of Space Fission Reactors

International Nuclear Information System (INIS)

Hixson, Laurie L.; Houts, Michael G.; Clement, Steven D.

2004-01-01

Performance and demonstration testing are critical to the success of a space fission reactor program. However, the type and extent to which testing of space reactors should be performed has been a point of discussion within the industry for many years. With regard to full power ground nuclear tests, questions such as 'Do the benefits outweigh the risks? Are there equivalent alternatives? Can a test facility be constructed (or modified) in a reasonable amount of time? Will the test article accurately represent the flight system? Are the costs too restrictive?' have been debated for decades. There are obvious benefits of full power ground nuclear testing such as obtaining systems integrated reliability data on a full-scale, complete end-to-end system. But these benefits come at some programmatic risk. In addition, this type of testing does not address safety related issues. This paper will discuss and assess these and other technical considerations essential in deciding which type of performance and demonstration testing to conduct on space fission reactor systems. (authors)

Design study on sodium-cooled middle-scale modular reactor

International Nuclear Information System (INIS)

Kisohara, Naoyuki; Hishida, Masahiko; Nibe, Nobuaki

2003-09-01

In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled middle-scale modular reactor, which has a possibility to fulfill the design requirements of the F/S. This report summarizes the results of the design study on the sodium-cooled middle-scale modular reactor performed in JFY2002, which is the second year of Phase 2. The construction cost of the sodium-cooled middle-scale modular reactor, which has been constructed in JFY2002, was almost achieved the economical goal. But its achievability was not sufficient to accept the concept. In order to reduce the construction cost, the plant concept has been re-constructed based on the 50 MWe plant studied in JFY2002. After that, fundamental specifications of main systems and components for the new concept have been set, and critical subjects have been examined and evaluated. In addition, in order to achieve the further cost reduction, the plant with simplified secondary system, the plant with electric magnetic pump in secondary system, and the fuel handling system are examined and evaluated. As a result of this study, the plant concept of the sodium-cooled middle-scale modular reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000 yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

JNDC nuclear data library of fission products

International Nuclear Information System (INIS)

Tasaka, Kanji; Ihara, Hitoshi; Akiyama, Masatsugu; Yoshida, Tadashi; Matumoto, Zyun-itiro; Nakasima, Ryuzo

1983-10-01

The JNDC (Japanese Nuclear Data Committee) FP (Fission Product) nuclear data library for 1172 fission products is described in this report. The gross theory of beta decay has been used extensively for estimating unknown decay data and also some of known decay data with poor accuracy. The calculated decay powers of fission products using the present library show excellent agreement with the latest measurements at ORNL (Oak Ridge National Laboratory), LANL (Los Alamos National Laboratory) and UTT (University of Tokyo, Tokai) for cooling times shorter than 10 3 s after irradiation. The calculated decay powers by the existing libraries showed systematic deviations at short cooling times; the calculated beta and gamma decay powers after burst fission were smaller than the experimental results for cooling times shorter than 10 s, and in the cooling time range 10 to 10 3 s the beta-decay power was larger than the measured values and the gamma decay power smaller than the measured results. The present JNDC FP nuclear data library resolved these discrepancies in the short cooling time ranges. The decay power of fission products has been calculated for ten fission types and the results have been fitted by an analytical function with 31 exponentials. This permits the easy application of the present results of decay power calculations to a LOCA (Loss-of-Coolant Accident) analysis of a light water reactor and so on. (author)

Power-balance analysis of muon-catalyzed fusion-fission hybrid reactor systems

International Nuclear Information System (INIS)

Miller, R.L.; Krakowski, R.A.

1985-01-01

A power-balance model of a muon-catalyzed fusion system in the context of a fission-fuel factory is developed and exercised to predict the required physics performance of systems competitive with either pure muon-catalyzed fusion systems or thermonuclear fusion-fission fuel factory hybrid systems

Method of fission product beta spectra measurements for predicting reactor anti-neutrino emission

Energy Technology Data Exchange (ETDEWEB)

Asner, D.M.; Burns, K.; Campbell, L.W.; Greenfield, B.; Kos, M.S., E-mail: markskos@gmail.com; Orrell, J.L.; Schram, M.; VanDevender, B.; Wood, L.S.; Wootan, D.W.

2015-03-11

The nuclear fission process that occurs in the core of nuclear reactors results in unstable, neutron-rich fission products that subsequently beta decay and emit electron antineutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to today's precision measurements of neutrino mixing angles. The prediction of the absolute flux and energy spectrum of the emitted reactor neutrinos hinges upon a series of seminal papers based on measurements performed in the 1970s and 1980s. The steadily improving reactor neutrino measurement techniques and recent reconsiderations of the agreement between the predicted and observed reactor neutrino flux motivates revisiting the underlying beta spectra measurements. A method is proposed to use an accelerator proton beam delivered to an engineered target to yield a neutron field tailored to reproduce the neutron energy spectrum present in the core of an operating nuclear reactor. Foils of the primary reactor fissionable isotopes placed in this tailored neutron flux will ultimately emit beta particles from the resultant fission products. Measurement of these beta particles in a time projection chamber with a perpendicular magnetic field provides a distinctive set of systematic considerations for comparison to the original seminal beta spectra measurements. Ancillary measurements such as gamma-ray emission and post-irradiation radiochemical analysis will further constrain the absolute normalization of beta emissions per fission. The requirements for unfolding the beta spectra measured with this method into a predicted reactor neutrino spectrum are explored.

Reactor auxiliary cooling facility and coolant supplying method therefor

International Nuclear Information System (INIS)

Ando, Koji; Kinoshita, Shoichiro.

1996-01-01

A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

Reactor auxiliary cooling facility and coolant supplying method therefor

Energy Technology Data Exchange (ETDEWEB)

Ando, Koji; Kinoshita, Shoichiro

1996-06-07

A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Verma, V., E-mail: vasudha.verma@physics.uu.se [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Barbot, L.; Filliatre, P. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Hellesen, C. [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Jammes, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Svärd, S. Jacobsson [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden)

2017-07-11

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment. - Highlights: • Studied possibility of using SPNDs as in-core detectors in SFRs. • Study done to detect local power profile changes when reactor is at nominal power. • SPND with a Pt-emitter gives measurable prompt current of the order of 600 nA/m. • Dominant proportion of prompt response is maintained throughout the operation. • Detector signal gives dynamic information on the power fluctuations.

Proceedings of 'workshop on Pb-alloy cooled fast reactor'

International Nuclear Information System (INIS)

Kim, Sang Ji; Kim, Yong Hee; Hong, Ser Gi

2003-06-01

The objective of 'Workshop on Pb-Alloy Cooled Fast Reactor', held in Taejeon, Korea on May 6, 2003, is to enhance the basic knowledge in this area by facilitating the exchange of information and discussions about problematic area of design aspects. There were five presentations from three different countries and about 25 participants gathered during the workshop. The topics covered in the workshop include benefits and drawbacks of Pb-alloy and Sodium coolant, two Pb-alloy cooled 900 MWt reactor designs using both B4C rods and NSTs, BREST-300 breakeven reactor and transmutation effectiveness of LLFPs in the typical thermal/fast neutron systems. The generic conclusion for the Pb-alloy cooled fast reactor from this workshop is as follows: 1) It has a potential to satisfy the goals established for the Generation-IV reactor concepts, so it has a bright future. 2) As a fast neutron system with a moderate breeding or a conversion, it is flexible in its roles and has superior safety characteristics over sodium coolant because of Pb-alloy's chemical inertness with water/air and high boiling temperature

Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

International Nuclear Information System (INIS)

Klein, Andrew; Lenhof, Renae; Deason, Wesley; Harter, Jackson

2015-01-01

Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

Energy Technology Data Exchange (ETDEWEB)

Klein, Andrew [Oregon State Univ., Corvallis, OR (United States). Nuclear Engineering and Radiation Health Physics; Matthews, Topher [Oregon State Univ., Corvallis, OR (United States); Lenhof, Renae [Oregon State Univ., Corvallis, OR (United States); Deason, Wesley [Oregon State Univ., Corvallis, OR (United States); Harter, Jackson [Oregon State Univ., Corvallis, OR (United States)

2015-01-16

Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

Conceptual study of a complementary scram system for liquid metal cooled nuclear reactors

International Nuclear Information System (INIS)

Vanmaercke, S.; Van den Eynde, G.; Tijskens, E.; Bartosiewicz, Y.

2009-01-01

GEN-IV reactors promise higher safety and reliability as one of the major improvements over previous generations of reactors. To achieve that, all GEN-IV reactor concepts require two completely independent shutdown systems that rely on different operating principles. For liquid metal cooled reactors the first system is an absorber-rod based solution. The second system that by requirement should rely on another principle, is however quite a challenge to design. The second system used in current PWR reactors is to dissolve a neutron absorber, boric acid, into the primary coolant. This method cannot be used in liquid metal cooled reactors because of the high cost of cleaning the coolant. In this paper an overview of the existing literature on scram systems is given, each with their advantages and limitations. A promising new concept is also presented. This concept leads to a totally passive self activating device using small absorbing particles that flow into a dedicated channel to shutdown the reactor. The system consists of tubes filled with particles of an absorber material. During normal operation, these particles are kept above the active core by means of a metallic seal. In case of an accident, the system is activated by the temperature increase in the coolant. This leads to melting of the metal seal. The ongoing work conducted at SCK·CEN and UCL/TERM aims at assessing the reliability of this new concept both experimentally and numerically. This study is multidisciplinary as neutronic and thermal hydraulics issues are tackled. Most challenging is however the thermal hydraulics related to understanding and predicting the liberation and flow of the absorber particles during a shutdown. Simple experiments are envisaged to compare to numerical simulations using the Discrete Element Method for simulating the particles. In a later stage this will be coupled with Smoothed Particles Hydrodynamics for simulating the melting of the seal. Some preliminary experimental and