A mechanism for proven technology foresight for emerging fast reactor designs and concepts

Energy Technology Data Exchange (ETDEWEB)

Anuar, Nuraslinda, E-mail: nuraslinda@uniten.edu.my; Muhamad Pauzi, Anas, E-mail: anas@uniten.edu.my [College of Engineering, Universiti Tenaga Nasional, Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia)

2016-01-22

The assessment of emerging nuclear fast reactor designs and concepts viability requires a combination of foresight methods. A mechanism that allows for the comparison and quantification of the possibility of being a proven technology in the future, β for the existing fast reactor designs and concepts is proposed as one of the quantitative foresight method. The methodology starts with the identification at the national or regional level, of the factors that would affect β. The factors are then categorized into several groups; economic, social and technology elements. Each of the elements is proposed to be mathematically modelled before all of the elemental models can be combined. Once the overall β model is obtained, the β{sub min} is determined to benchmark the acceptance as a candidate design or concept. The β values for all the available designs and concepts are then determined and compared with the β{sub min}, resulting in a list of candidate designs that possess the β value that is larger than the β{sub min}. The proposed methodology can also be applied to purposes other than technological foresight.

Design concepts and status of the Korean next generation reactor (KNGR)

International Nuclear Information System (INIS)

Cho, Sung Jae; Kim, Han Gon

1999-01-01

The national project to develop KNGR, a 4000 MWth evolutionary advanced light water reactor (ALWR), has been organized in three phases according to the development status in 1992. During the first phase, the top-tier design requirements and the design concepts to meet the requirements had been established. The project is currently in the second phase of which the major objective is to complete the basic design sufficient to confirm the plant safety. This paper describes the overall design concepts and status of the KNGR briefly which developed and/or being developed through the project. (author)

Fuel transfer cask concept design for reactor TRIGA PUSPATI (RTP)

International Nuclear Information System (INIS)

Ahmad Nabil Ab Rahim; Phongsakorn Prak; Tonny Lanyau; Mohd Fazli Zakaria

2010-01-01

Reactor Triga PUSPATI (RTP) has been operated since 1982 till now. For such long period, the organization feels the need to upgrade the power from 1 MW to 3 MW which involved changing new fuels. Spent fuels will be stored in a Spent Fuel Pool. The process of transferring spent fuels into Spent Fuels Pool required a fuel transfer cask. This paper discussed the design concept for the fuel transfer cast which is essential equipment for reactor upgrading mission. (author)

Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

International Nuclear Information System (INIS)

Yun, Sunghwan; Kim, Sang Ji

2015-01-01

In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH 2 and B 4 C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Generation-IV nuclear reactors, SFR concept

International Nuclear Information System (INIS)

Dufour, P.

2010-01-01

In this presentation author deals with development of sodium-cooled fast reactors and lead-cooled fast reactors. He concluded that: - SFR is a proved concept that has never achieved industrial deployment; - The GEN IV objectives need to reconsider the design of both the core and the reactor design : innovations are being analysed; Future design will benefit from considerable feedback of design, licensing, construction and operation of PX, SPX, etc.

Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Working Material

International Nuclear Information System (INIS)

2013-01-01

The objective of the TM on “Liquid metal reactor concept: core design and structural materials” was to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials. Main results arising from national and international R&D programmes and projects in the field were reviewed, and new activities to be carried out under the IAEA aegis were identified on the basis of the analysis of current research and technology gaps

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

International Nuclear Information System (INIS)

Allen, Francis; Bonin, Hugues

2008-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

SEBREZ: an inertial-fusion-reactor concept

International Nuclear Information System (INIS)

Meier, W.R.

1982-01-01

The neutronic aspects of an inertial fusion reactor concept that relies on asymmetrical neutronic effects to enhance the tritium production in the breeding zones have been studied. We find that it is possible to obtain a tritium breeding ratio greater than 1.0 with a chamber configuration in which the breeding zones subtend only a fraction of the total solid angle. This is the origin of the name SEBREZ which stands for SEgregated BREeding Zones. It should be emphasized that this is not a reactor design study; rather this study illustrates certain neutronic effects in the context of a particular reactor concept. An understanding of these effects forms the basis of a design technique which has broader application than just the SEBREZ concept

The design features and safety concepts of the nuclear heating reactor developed in China

International Nuclear Information System (INIS)

Zheng Wenxiang; Wang Dazhong

1995-01-01

Based on the specific conditions of the nuclear heat applications and the development objectives of the advanced reactors, the nuclear heating reactor (NHR) exploited in China has adhered to the new safety concepts and been designed with a number of advanced features, including the integrated arrangement, full power natural circulation capacity, self-pressurized performance, dynamically-hydraulic control rod drive and passive safety systems, so that higher standard of safety as well as simplification in the plant systems and improvement in economic viability has been achieved. This paper describes the special consideration in the design as well as the main design features and safety concepts of the NHR. Some experimental and analytical results are also presented to demonstrate the NHR safety features

Small propulsion reactor design based on particle bed reactor concept

International Nuclear Information System (INIS)

Ludewig, H.; Lazareth, O.; Mughabghab, S.; Perkins, K.; Powell, J.R.

1989-01-01

In this paper Particle Bed Reactor (PBR) designs are discussed which use 233 U and /sup 242m/Am as fissile materials. A constant total power of 100MW is assumed for all reactors in this study. Three broad aspects of these reactors is discussed. First, possible reactor designs are developed, second physics calculations are outlined and discussed and third mass estimates of the various candidates reactors are made. It is concluded that reactors with a specific mass of 1 kg/MW can be envisioned of 233 U is used and approximately a quarter of this value can be achieved if /sup 242m/Am is used. If this power level is increased by increasing the power density lower specific mass values are achievable. The limit will be determined by uncertainties in the thermal-hydraulic analysis. 5 refs., 5 figs., 6 tabs

Review of the current status of linear hybrid reactor concepts

International Nuclear Information System (INIS)

Schultz, K.R.

1977-07-01

A review was made of the current status of linear fusion-fission hybrid reactor design studies in the USA. The linear hybrid reactor concepts reviewed include the linear theta-pinch hybrid reactor being studied at Los Alamos Scientific Laboratory, the electron beam-heated solenoid hybrid reactor under development at Physics International Co., the laser-heated solenoid hybrid reactor being investigated at Mathematical Sciences Northwest, Inc., and the linear fusion waste burning reactor being studied at General Atomic Company. The discussion addresses confinement and heating mechanisms for each concept, as well as the hybrid blanket designs. The current state of the four reactor designs is summarized and the performance of the various concepts compared

Design of subjects training on reactor simulator and feasibility study - toward the empirical evaluation of interface design concept

International Nuclear Information System (INIS)

Yamaguchi, Y.; Furukawa, H.; Tanabe, F.

1998-01-01

On-going JAERI's project for empirical evaluation of the ecological interface design concept was first described. The empirical evaluation is planned to be proceeded through three consecutive steps; designing and actual implementation of the interface on reactor simulator, verification of the interface created, and the validation by the simulator experiment. For conducting the project, three different experimental resources are prerequisite, that are, data analysis method for identifying the operator's strategies, experimental facility including reactor simulator, and experimental subjects or subjects training method. Among the three experimental resources, subjects training method was recently designed and a simulator experiment was earned out in order to examine the feasibility of the designed training method. From the experiment and analysis of the experimental records, we could conclude that it is feasible that the experimental subjects having an appropriate technical basis can gain the sufficient competence for evaluation work of the interface design concept by using the training method designed. (author)

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

International Nuclear Information System (INIS)

Hodzic, A.; Kugeler, K.

1997-01-01

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

Trends and developments in magnetic confinement fusion reactor concepts

International Nuclear Information System (INIS)

Baker, C.C.; Carlson, G.A.; Krakowski, R.A.

1981-01-01

An overview is presented of recent design trends and developments in reactor concepts for magnetic confinement fusion. The paper emphasizes the engineering and technology considerations of commercial fusion reactor concepts. Emphasis is placed on reactors that operate on the deuterium/tritium/lithium fuel cycle. Recent developments in tokamak, mirror, and Elmo Bumpy Torus reactor concepts are described, as well as a survey of recent developments on a wide variety of alternate magnetic fusion reactor concepts. The paper emphasizes recent developments of these concepts within the last two to three years

Light ion driven inertial fusion reactor concepts

International Nuclear Information System (INIS)

Cook, D.L.; Sweeney, M.A.; Buttram, M.T.; Prestwich, K.R.; Moses, G.A.; peterson, R.R.; Lovell, E.G.; Englestad, R.L.

1980-01-01

The possibility of designing fusion reactor systems using intense beams of light ions has been investigated. concepts for beam production, transport, and focusing on target have been analyzed in light of more conservative target performance estimates. Analyses of the major criteria which govern the design of the beam-target-cavity tried indicate the feasibility of designing power systems at the few hundred megawatt (electric) level. This paper discusses light ion fusion reactor (LIFR) concepts and presents an assessment of the design limitations through quantitative examples

New reactor concepts

International Nuclear Information System (INIS)

Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A.

1998-11-01

The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

Integral fast reactor concept inherent safety features

International Nuclear Information System (INIS)

Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

1987-01-01

The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFT development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: 1) a liquid metal (sodium) coolant, 2) a pool-type reactor primary system configuration, 3) an advanced ternary alloy metallic fuel, and 4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

Integral Fast Reactor concept inherent safety features

International Nuclear Information System (INIS)

Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

1986-01-01

The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: (1) a liquid metal (sodium) coolant, (2) a pool-type reactor primary system configuration, (3) an advanced ternary alloy metallic fuel, and (4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

Integral design concepts of advanced water cooled reactors. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1997-11-01

Under the sub-programme on non-electrical applications of advanced reactors, the International Atomic Energy Agency has been providing a worldwide forum for exchange of information on integral reactor concepts. Two Technical Committee meetings were held in 1994 and 1995 on the subject where state-of-the-art developments were presented. Efforts are continuing for the development of advanced nuclear reactors of both evolutionary and innovative design, for electricity, co-generation and heat applications. While single purpose reactors for electricity generation may require small and medium sizes under certain conditions, reactors for heat applications and co-generation would be necessary in the small and medium range and need to be located closer to the load centres. The integral design approach to the development of advanced light water reactors has received special attention over the past few years. Several designs are in the detailed design stage, some are under construction, one prototype is in operation. A need has been felt for guidance on a number of issues, ranging from design objectives to the assessment methodology needed to show how integral designs can meet these objectives, and also to identify their advantages and problem areas. The technical document addresses the current status of the design, safety and operational issues of integral reactors and recommends areas for future development

Analysis and evaluation of the Dual Fluid Reactor concept

Energy Technology Data Exchange (ETDEWEB)

Wang, Xiang

2017-06-27

The Dual Fluid Reactor is a molten salt fast reactor developed by IFK in Berlin based on the Gen-IV Molten-Salt Reactor concept and the Liquid-Metal Cooled Reactor. The design aims to combine these two concepts to improve these two concepts. The Dissertation focuses on the concept and performs diverse calculations and estimations on the subjects of neutron physics, depletion and thermal-hydraulic behaviors to validate the new features of the concept. Based on the results it is concluded that this concept is feasible to its desired purpose and with great potential.

Developing the MAPLE materials test reactor concept

International Nuclear Information System (INIS)

Lee, A.G.; Lidstone, R.F.; Donnelly, J.V.

1992-05-01

MAPLE-MTR is a new multipurpose research facility being planned by AECL Research as a possible replacement for the 35-year-old NRU reactor. In developing the MAPLE-MTR concept, AECL is starting from the recent design and licensing experience with the MAPLE-X10 reactor. By starting from technology developed to support the MAPLE-X10 design and adapting it to produce a concept that satisfies the requirements of fuel channel materials testing and fuel irradiation programs, AECL expects to minimize the need for major advances in nuclear technology (e.g., fuel, heat transfer). Formulation of the MAPLE-MTR concept is at an early stage. This report describes the irradiation requirements of the research areas, how these needs are translated into design criteria for the project and elements of the preliminary design concept

Assessment of nuclear reactor concepts for low power space applications

Science.gov (United States)

Klein, Andrew C.; Gedeon, Stephen R.; Morey, Dennis C.

1988-01-01

The results of a preliminary small reactor concepts feasibility and safety evaluation designed to provide a first order validation of the nuclear feasibility and safety of six small reactor concepts are given. These small reactor concepts have potential space applications for missions in the 1 to 20 kWe power output range. It was concluded that low power concepts are available from the U.S. nuclear industry that have the potential for meeting both the operational and launch safety space mission requirements. However, each design has its uncertainties, and further work is required. The reactor concepts must be mated to a power conversion technology that can offer safe and reliable operation.

Basic concept of fuel safety design and assessment for sodium-cooled fast reactor

International Nuclear Information System (INIS)

Nakae, Nobuo; Baba, Toshikazu; Kamimura, Katsuichiro

2013-03-01

'Philosophy in Safety Evaluation of Fast Breeder Reactors' was published as a guideline for safety design and safety evaluation of Sodium-Cooled Fast Reactor in Japan. This guideline points out that cladding creep and swelling due to internal pressure should be taken into account since the fuel is used under high temperature and high burnup, and that fuel assembly deformation and the prevention from coolant channel blockage should be taken into account in viewpoints of nuclear and thermal hydraulic design. However, the requirements including their criteria and evaluation items are not described. Two other domestic guidelines related to core design are applied for fuel design of fast reactor, but the description is considered to not be enough to practically use. In addition, technical standard for nuclear fuel used in power reactors is also applied for fuel inspection. Therefore, the technical standard and guideline for fuel design and safety evaluation are considered to be very important issue for nuclear safety regulation. This document has been developed according to the following steps: The guidelines and the technical standards, which are prepared in foreign countries and international organization, were reviewed. The technical background concerning fuel design and safety evaluation for fast reactor was collected and summarized in the world wide scale. The basic concept of fuel safety design and assessment for sodium-cooled fast reactor was developed by considering a wide range of views of the specialists in Japan. In order to discuss the content with foreign specialists IAEA Consultancy Meetings have been held on January, 2011 and January, 2012. The participants of the meeting came from USA, UK, EC, India, China and South Korea. The specialists of IAEA and JNES were also joined. Although this document is prepared for application to 'Monju'(prototype LMFR), it may be applied to experimental, demonstration and commercial types of LMFR after revising it by taking

Penn State advanced light water reactor concept

International Nuclear Information System (INIS)

Borkowski, J.A.; Smith, K.A.; Edwards, R.M.; Robinson, G.E.; Schultz, M.A.; Klevans, E.H.

1987-01-01

The accident at Three Mile Island heightened concerns over the safety of nuclear power. In response to these concerns, a research group at the Pennsylvania State University (Penn State) undertook the conceptual design of an advanced light water reactor (ALWR) under sponsorship of the US Dept. of Energy (DOE). The design builds on the literally hundreds of years worth of experience with light water reactor technology. The concept is a reconfigured pressurized water reactor (PWR) with the capability of being shut down to a safe condition simply by removing all ac power, both off-site and on-site. Using additional passively activated heat sinks and replacing the pressurizer with a pressurizing pump system, the concept essentially eliminates the concerns of core damage associated with a total station blackout. Evaluation of the Penn State ALWR concept has been conducted using the EPRI Modular Modeling System (MMS). Results show that a superior response to normal operating transients can be achieved in comparison to the response with a conventional PWR pressurizer. The DOE-sponsored Penn State ALWR concept has evolved into a significant reconfiguration of a PWR leading to enhanced safety characteristics. The reconfiguration has touched a number of areas in overall plant design including a shutdown turbine in the secondary system, additional passively activated heat sinks, a unique primary side pressurizing concept, a low pressure cleanup system, reactor building layout, and a low power density core design

Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

Directory of Open Access Journals (Sweden)

Yeong-il Kim

2013-01-01

Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

Design of a New Research Reactor: Preliminary Conceptual Design (3rd Year)

International Nuclear Information System (INIS)

Park, Cheol; Lee, B. C.; Chae, H. T. and others

2006-01-01

A research reactor design is a kind of integral engineering project and a process to obtain a concrete shape through several years of concept development, conceptual design, basic design and detail design. So it requires close cooperation in various areas as well as lots of manpower and cost. The overall process at each stage may be said to be similar except for some stage-specific works. In 2005 as last year of a concept development stage, investigations on the various concepts of the fuel, reactor structure and systems which can meet the requirements established. The requirements for the process systems and I and C systems have also been embodied. The major tasks planned at the early of 2005 have been performed for each area of reactor design as follows: Establishment of the fuel and reactor core concept, and the core analysis, Preliminary thermal-hydraulic and safety analyses for the conceptual cores, Establishment and improvement of analysis system, Concept developments of the reactor structures and major systems, Test and test plan to verify the developed concepts, International cooperation to establish the foundations for exporting a research reactor

Inertial fusion reactor designs

International Nuclear Information System (INIS)

Meier, W.

1987-01-01

In this paper, a variety of reactor concepts are proposed. One of the prime concerns is dealing with the x-rays and debris that are emitted by the target. Internal neutron shielding can reduce radiation damage and activation, leading to longer life systems, reduced activation and fewer safety concerns. There is really no consensus on what the best reactor concept is at this point. There has been virtually no chamber technology development to date. This is the flip side of the coin of the separability of the target physics and the reactor design. Since reactor technology has not been required to do target experiments, it's not being developed. Economic analysis of conceptual designs indicates that ICF can be economically competitive with magnetic fusion, fission and fossil plants

An innovative fuel design concept for improved Light Water Reactor performance and safety

International Nuclear Information System (INIS)

Tulenko, J.S.; Connell, R.G.

1993-01-01

The primary goal of this research is to develop a new fuel design which will have improved thermal/mechanical performance characteristics greatly superior to current thermal and mechanical design performance. The mechanical/thermal constraints define the lifetime of the fuel, the maximum power at which the fuel can be operated, the probability of fuel failure over core lifetime, and the integrity of a core during a transient excursion. The thermal/mechanical limits act to degrade fuel integrity when they are violated. The purpose of this project is to investigate a novel design for light water reactor fuel which will extend fuel performance limits and improve reactor safety even further than is currently achieved. This project is investigating liquid metal bonding of LWR fuel in order to radically decrease fuel centerline temperatures which has major performance and safety benefits. The project will verify the compatibility of the liquid metal bond with both the fuel pellets and cladding material, verify the performance enhancement features of the new design over the fuel lifetime, and verify the economic fabricability of the concept and will show how this concept will benefit the LWR nuclear industry

Rapid-L Operator-Free Fast Reactor Concept Without Any Control Rods

International Nuclear Information System (INIS)

Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

2003-01-01

The 200-kW(electric) uranium-nitride-fueled lithium-cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for a lunar base power system. It is one of the variants of the RAPID (Refueling by All Pins Integrated Design) fast reactor concept, which enables quick and simplified refueling. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 2700 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 yr.Unique challenges in reactivity control systems design have been addressed in the RAPID-L concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt 6 Li as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs, and LRMs, RAPID-L can be operated without an operator. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, the RAPID-L reactor concept and its transient characteristics are presented

A computer code for Tokamak reactor concepts evaluation

International Nuclear Information System (INIS)

Rosatelli, F.; Raia, G.

1985-01-01

A computer package has been developed which could preliminarily investigate the engineering configuration of a tokamak reactor concept. The code is essentially intended to synthesize, starting from a set of geometrical and plasma physics parameters and the required performances and objectives, three fundamental components of a tokamak reactor core: blanket+shield, TF magnet, PF magnet. An iterative evaluation of the size, power supply and cooling system requirements of these components allows the judgment and the preliminary design optimization on the considered reactor concept. The versatility of the code allows its application both to next generation tokamak devices and power reactor concepts

PIUS principle and the SECURE reactor concepts

International Nuclear Information System (INIS)

Hannerz, K.

1987-01-01

The author introduces the SECURE reactor concept, a reactor intended for producing heat for district heating grids, desalination, and certain process industries. A detailed design of a 400 MWth plant has been completed and is being offered commercially. The authors present first, a summary of the current situation and then the design philosophy of the SECURE reactor concepts. The authors propose a design based on a light water reactor, as opposed to high temperature gas cooled reactor, but introduce new features which are designed to eliminate the element of human error in preparing for and handling emergencies. The authors propose two rules to avoid overheating, i.e.., the PIUS design principle, which are: to keep the core submerged in water; and to ensure that the rate of heat generation in the submerged core is low enough to avoid overheating of the fuel (dryout). The acronym PIUS stands for Process Inherent Ultimate Safety. A detailed system modeling is given of the PIUS primary system. The design of the plant is divided into two parts: the nuclear island, which is comprised of the concrete vessel and its contents; and the balance of the plant, which is comprised of all other components, including the turbine plant

Concept and designs of new-generation fast reactors

International Nuclear Information System (INIS)

Mitenkov, F.M.

1993-01-01

This article discusses the general safety requirements and characteristics for future nuclear power plants. It examines various designs - loop, block, and integrated layouts for reactors. Specifically, the article focuses an integrated design for sodium-cooled fast reactors noting that the BN-600 reactor has operated accident-free over the past 12 years. An obvious advantage of this scheme is that the coolant of the primary loop is localized in one volume (in a vessel), there are no short connections and large-diameter pipes, which of course sharply reduces the probability in coolant leaks. With an integrated scheme the problem of embrittlement of the reactor vessel by neutron irradiation is obviated. The neutron fluence for the vessels of the AST-500 and VPBER-600 reactors, built with an integrated scheme, is less than 10 17 cm -2 . Such a fluence does not cause any appreciable change in the mechanical properties of the vessel steel. The integrated layout of the reactor makes it possible to build a containment vessel. In this case it is possible to eliminate the danger of the reactor core drying out and thus cooling of the reactor in emergency situations can be simplified substantially. In an integrated layout, however, access is more difficult to the equipment inside the reactor, thus limiting or complicating maintenance work. The integrated layout, therefore, requires the use of highly reliable equipment built according to designs that have been proven in operation and have been passed representative service-life tests under laboratory conditions. The integrated layout considerably increases the mass and size characteristics of the reactor. New solutions thus are needed for the organization of work on reactor fabrication and assembly. In the case of the BN-600 and Superphenix reactors the welding of the reactor vessels and the assembly work were done on the building site

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

2003-01-01

Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-03-01

A conceptual design study (option C) has been carried out for the fusion experimental reactor (FER). In addition to design of the tokamak reactor and associated systems based on the reference design specifications, feasibility of a water-shield reactor concept was examined as a topical study. The design study for the reference tokamak reactor has produced a reactor concept for the FER, along with major R D items for the concept, based on close examinations on thermal design, electromagnetics, neutronics and remote maintenance. Particular efforts have been directed to the area of electromagnetics. Detailed analyses with close simulation models have been performed on PF coil arrangements and configurations, shell effects of the blanket for plasma position unstability, feedback control, and eddy currents during disruptions. The major design specifications are as follows; Peak fusion power 437 MW Major radius 5.5 m Minor radius 1.1 m Plasma elongation 1.5 Plasma current 5.3 MA Toroidal beta 4 % Field on axis 5.7 T (author)

Safety characteristics of the integral fast reactor concept

International Nuclear Information System (INIS)

Marchaterre, J.F.; Cahalan, J.E.; Sevy, R.H.; Wright, A.E.

1985-01-01

The Integral Fast Reactor (IFR) concept is an innovative approach to liquid metal reactor design which is being studied by Argonne National Laboratory. Two of the key features of the IFR design are a metal fuel core design, based on the fuel technology developed at EBR-II, and an integral fuel cycle with a colocated fuel cycle facility based on the compact and simplified process steps made possible by the use of metal fuel. The paper presents the safety characteristics of the IFR concept which derive from the use of metal fuel. Liquid metal reactors, because of the low pressure coolant operating far below its boiling point, the natural circulation capability, and high system heat capacities, possess a high degree of inherent safety. The use of metallic fuel allows the reactor designer to further enhance the system capability for passive accommodation of postulated accidents

Overview of in-vessel retention concept involving level of passivity: with application to evolutionary pressurized water reactor design

International Nuclear Information System (INIS)

Ghyym, Seong H.

1998-01-01

In this work, one strategy of severe accident management, the applicability of the in-vessel retention (IVR) concept, which has been incorporated in passive type reactor designs, to evolutionary type reactor designs, is examined with emphasis on the method of external reactor vessel cooling (ERVC) to realize the IVR concept in view of two aspects: for the regulatory aspect, it is addressed in the context of the resolution of the issue of corium coolability; for the technical one, the reliance on and the effectiveness of the IVR concept are mentioned. Additionally, for the ERVC method to be better applied to designs of the evolutionary type reactor, the conditions to be met are pointed out in view of the technical aspect. Concerning the issue of corium coolability/quenchability, based on results of the review, plausible alternative strategies are proposed. According to the decision maker's risk behavior, these would help materialize the conceptual design for evolutionary type reactors, especially Korea Next Generation Reactors (KNGRs), which have been developing at the Korea Electric Power Research Institute (KEPRI): (A1) Strategy 1A: strategy based on the global approach using the reliance on the wet cavity method; (A2) Strategy 1B: strategy based on the combined approach using both the reliance on the wet cavity method and the counter-measures for preserving containment integrity; (A3) Strategy 2A: strategy based on the global approach to the reliance on the ERVC method; (A4) Strategy 2B: strategy based on the balanced approach using both the reliance on the ERVC method and the countermeasures for preserving containment integrity. Finally, in application to an advanced pressurized water reactor (PWR) design, several recommendations are made in focusing on both monitoring the status of approaches and preparing countermeasures in regard to the regulatory and the technical aspects

The advanced MAPLE reactor concept

International Nuclear Information System (INIS)

Lidstone, R.F.; Lee, A.G.; Gillespie, G.E.; Smith, H.J.

1989-01-01

In Canada the need for advanced neutron sources has long been recognized. During the past several years Atomic Energy of Canada Limited (AECL) has been developing the new MAPLE multipurpose reactor concept. To date, the MAPLE program has focused on the development of a modest-cost multipurpose medium-flux neutron source to meet contemporary requirements for applied and basic research using neutron beams, for small-scale materials testing and analysis and for radioisotope production. The basic MAPLE concept incorporates a compact light-water cooled and moderated core within a heavy water primary reflector to generate strong neutron flux levels in a variety of irradiation facilities. In view of renewed Canadian interest in a high-flux neutron source, the MAPLE group has begun to explore advanced concepts based on AECL's experience with heavy water reactors. The overall objective is to define a high-flux facility that will support materials testing for advanced power reactors, new developments in extracted neutron-beam applications, and/or production of radioisotopes. The design target is to attain performance levels of HFR-Grenoble, HFBR, HFIR in a new heavy water-cooled, -moderated,-reflected reactor based on rodded LEU fuel. Physics, shielding, and thermohydraulic studies have been performed for the MAPLE heavy water reactor. 14 refs., 4 figs., 1 tab

A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

Energy Technology Data Exchange (ETDEWEB)

Jacques V Hugo; David I Gertman; Jeffrey C Joe

2014-08-01

This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

PRISM [Power Reactor Inherently Safe Module] design concept enhances waste management

International Nuclear Information System (INIS)

Thompson, M.L.; Berglund, R.C.

1989-01-01

PRISM, a modular advanced liquid metal reactor (ALMR), has been designed conceptually by GE under the US Department of Energy sponsorship. The concept design and analyses have been primarily focused on passive safety and improved construction and operating costs. Significantly, the unique design of multiple modules and features of PRISM enhance waste management over conventional reactor systems. This paper provides an overview of PRISM of these enhancements. Inherent to the ALMR's, the sodium coolant precludes crud buildup on reactor surfaces and in components and waste for disposal. Preliminary evaluations indicate this fundamental feature results in factors of 2-4 less waste volume and 2-3 orders of magnitude less curies per megawatt-electric for ultimate disposal. For example, the tap designed for sodium cleanup is expected to be exchanged only once every thirty years. Also, inherent to ALMR's, burning waste actinides and selected fission products to preclude their accumulation and burial is very attractive. The hard neutron spectrum of ALMR burns the actinides efficiently and is not poisoned by the actinides and fission products. The modular design of PRISM components (and the fuel cycle equipment) permit replacement without expensive and potentially hazardous volume reduction. For example, the functional components of the reference electromagnetic pump and IHK can be removed intact for waste disposal. Although development of the reference metal fuel is not completed, it is estimated that (low-level) waste from recycle of the fuel will result in significantly less volume than would be generated by aqueous recycle of oxide fuel. 6 refs., 10 figs

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)

PSA in design of passive/active safety reactors

International Nuclear Information System (INIS)

Sato, T.; Tanabe, A.; Kondo, S.

1995-01-01

PSAs in the design of advanced reactors are applied mainly in level 1 PSA areas. However, even in level 1 PSA, there are certain areas where special care must be taken depending on plant design concepts. This paper identifies these areas both for passive and active safety reactor concepts. For example, 'long-term PSA' and shutdown PSA are very important for a passive safety reactor concept from the standpoint of effectiveness of a grace period and passive safety systems. External events are also important for an active safety reactor concept. These kinds of special PSAs are difficult to conduct precisely in a conceptual design stage. This paper shows methods of conducting these kinds of special PSAs simply and conveniently and the use of acquired insights for the design of advanced reactors. This paper also clarifies the meaning or definition of a grace period from the standpoint of PSA

Advances in laser solenoid fusion reactor design

International Nuclear Information System (INIS)

Steinhauer, L.C.; Quimby, D.C.

1978-01-01

The laser solenoid is an alternate fusion concept based on a laser-heated magnetically-confined plasma column. The reactor concept has evolved in several systems studies over the last five years. We describe recent advances in the plasma physics and technology of laser-plasma coupling. The technology advances include progress on first walls, inner magnet design, confinement module design, and reactor maintenance. We also describe a new generation of laser solenoid fusion and fusion-fission reactor designs

Development of physical conceptions of fast reactors

International Nuclear Information System (INIS)

Khomyakov, Yu.S.; Matveev, V.I.; Moiseev, A.V.

2013-01-01

• Russian experience in developing fast reactors has proved clearly scientific justification of conceptual physical principles and their technical feasibility. • However, the potential of fast reactors caused by their physical features has not been fully realized. • In order to assure the real possibility of transition to the nuclear power with fast reactors by about 2030 it is necessary to consistently update fast reactor designs for solving the following key problems: - increasing of self-protection level of reactor core; - improvement of technical and economical characteristics; - solution of the problems related to the fuel supply of nuclear power and assimilation of closed nuclear fuel cycle; - disposal of long lived radioactive waste and transmutation of minor actinides. • Russian program (2010-2020) on the development of basic concepts of the new generation reactors implies successive solution of the above problems. • New technical decisions will be demonstrated by development and assimilation of the new reactors: - BN-800 – development of the fuel cycle infrastructure and mastering of the new types of fuel; - BN-1200 reactor – demonstration economical efficiency of fast reactor and new level of safety; - BREST development and demonstration new heavy liquid metal coolant technology and alternative design concept

In-vessel maintenance concepts for tokamak fusion reactors

International Nuclear Information System (INIS)

Kelly, V.P.; Berger, J.D.; Yount, J.A.

1983-01-01

Concepts for rail-mounted and guided in-vessel handling machines (IVM) for remote maintenance inside tokamak fusion reactors are described. The IVM designs are based on concepts for tethered remotely operated vehicles and feature the use of multiple manipulator arms for remote handling and remote-controlled TV cameras for remote viewing. The concepts include IVMs for both single or dual rail systems located in the top or bottom of the reactor vessel

Fusion reactor design studies: standard accounts for cost estimates

International Nuclear Information System (INIS)

Schulte, S.C.; Willke, T.L.; Young, J.R.

1978-05-01

The fusion reactor design studies--standard accounts for cost estimates provides a common format from which to assess the economic character of magnetically confined fusion reactor design concepts. The format will aid designers in the preparation of design concept costs estimates and also provide policymakers with a tool to assist in appraising which design concept may be economically promising. The format sets forth a categorization and accounting procedure to be used when estimating fusion reactor busbar energy cost that can be easily and consistently applied. Reasons for developing the procedure, explanations of the procedure, justifications for assumptions made in the procedure, and the applicability of the procedure are described in this document. Adherence to the format when evaluating prospective fusion reactor design concepts will result in the identification of the more promising design concepts thus enabling the fusion power alternatives with better economic potential to be quickly and efficiently developed

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

International Nuclear Information System (INIS)

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

1982-05-01

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

A breed and burn reshuffling scheme for an Astrid-like reactor concept design

Energy Technology Data Exchange (ETDEWEB)

Garcia C, E. Y.; Francois L, J. L., E-mail: eliasgarcerv@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

2016-09-15

The greenhouse emissions has a serious impact on environmental terms, reason why energy supply needs to be provided by low carbon emission technologies. Nuclear power is and environmentally friendly energy with future concepts and designs that ensure safety, and in this paper is taken as a solution to be considered in the energy supply mix. This study presents an extension on the operation of the Advanced Sodium Technological Reactor for Industrial demonstration (Astrid) nuclear reactor through a breed and burn operation to extend the operation of the reactor. The Astrid nuclear reactor is a fourth generation sodium-cooled fast reactor of 1500 MWt h, and it considers an innovative design: the low void effect core (Cfv: Coeur a Faible effet de Vidange sodium) due to the reactor configuration and the radial and axial position of the fuel subassemblies. Previous research in the Astrid-like cores aimed the model validation of a conventional oxide-fueled core and its comparison with a proposed metallic-fueled core. Taking into account the amount of fissile material (mainly 239-Pu) after the first cycle, a reshuffling scheme was suggested, which consists in changing strategically the position of the nuclear fuel assemblies when the reactivity drops near to the critical state of the reactor. Two different reshuffling schemes were simulated in every developed model, having operation extensions of 805 days and 1775 days for the oxide and metallic designs respectively. The implementation of the reshuffling schemes in the developed models enhanced the fuel utilization and could save up to 2.20 and 5.96 tons of plutonium for oxide and metallic designs respectively, which has an economic impact. The breeding of 239-Pu achieved in the fertile zone of the metallic design reached half of the initial concentration of the 239-Pu in the fissile zone and for the oxide design, the breeding reached one third of the initial concentration of the 239-Pu in the fissile zone. (Author)

A breed and burn reshuffling scheme for an Astrid-like reactor concept design

International Nuclear Information System (INIS)

Garcia C, E. Y.; Francois L, J. L.

2016-09-01

The greenhouse emissions has a serious impact on environmental terms, reason why energy supply needs to be provided by low carbon emission technologies. Nuclear power is and environmentally friendly energy with future concepts and designs that ensure safety, and in this paper is taken as a solution to be considered in the energy supply mix. This study presents an extension on the operation of the Advanced Sodium Technological Reactor for Industrial demonstration (Astrid) nuclear reactor through a breed and burn operation to extend the operation of the reactor. The Astrid nuclear reactor is a fourth generation sodium-cooled fast reactor of 1500 MWt h, and it considers an innovative design: the low void effect core (Cfv: Coeur a Faible effet de Vidange sodium) due to the reactor configuration and the radial and axial position of the fuel subassemblies. Previous research in the Astrid-like cores aimed the model validation of a conventional oxide-fueled core and its comparison with a proposed metallic-fueled core. Taking into account the amount of fissile material (mainly 239-Pu) after the first cycle, a reshuffling scheme was suggested, which consists in changing strategically the position of the nuclear fuel assemblies when the reactivity drops near to the critical state of the reactor. Two different reshuffling schemes were simulated in every developed model, having operation extensions of 805 days and 1775 days for the oxide and metallic designs respectively. The implementation of the reshuffling schemes in the developed models enhanced the fuel utilization and could save up to 2.20 and 5.96 tons of plutonium for oxide and metallic designs respectively, which has an economic impact. The breeding of 239-Pu achieved in the fertile zone of the metallic design reached half of the initial concentration of the 239-Pu in the fissile zone and for the oxide design, the breeding reached one third of the initial concentration of the 239-Pu in the fissile zone. (Author)

Transient Analysis Needs for Generation IV Reactor Concepts

International Nuclear Information System (INIS)

Siefken, L.J.; Harvego, E.A.; Coryell, E.W.; Davis, C.B.

2002-01-01

The importance of nuclear energy as a vital and strategic resource in the U. S. and world's energy supply mix has led to an initiative, termed Generation IV by the U.S. Department of Energy (DOE), to develop and demonstrate new and improved reactor technologies. These new Generation IV reactor concepts are expected to be substantially improved over the current generation of reactors with respect to economics, safety, proliferation resistance and waste characteristics. Although a number of light water reactor concepts have been proposed as Generation IV candidates, the majority of proposed designs have fundamentally different characteristics than the current generation of commercial LWRs operating in the U.S. and other countries. This paper presents the results of a review of these new reactor technologies and defines the transient analyses required to support the evaluation and future development of the Generation IV concepts. The ultimate objective of this work is to identify and develop new capabilities needed by INEEL to support DOE's Generation IV initiative. In particular, the focus of this study is on needed extensions or enhancements to SCDAP/RELAP5/3D code. This code and the RELAP5-3D code from which it evolved are the primary analysis tools used by the INEEL and others for the analysis of design-basis and beyond-design-basis accidents in current generation light water reactors. (authors)

Design of an organic simplified nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-15

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

Design of an Organic Simplified Nuclear Reactor

Directory of Open Access Journals (Sweden)

Koroush Shirvan

2016-08-01

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

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)

Modular Stellarator Fusion Reactor (MSR) concept

International Nuclear Information System (INIS)

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

1981-01-01

A preliminary conceptual study has been made of the Modulator Stellarator Reactor (MSR) as a stedy-state, ignited, DT-fueled, magnetic fusion reactor. The MSR concept combines the physics of classic stellarator confinement with an innovative, modular-coil design. Parametric tradeoff calculations are described, leading to the selection of an interim design point for a 4.8-GWt plant based on Alcator transport scaling and an average beta value of 0.04 in an l = 2 system with a plasma aspect ratio of 11. Neither an economic analysis nor a detailed conceptual engineering design is presented here, as the primary intent of this scoping study is the elucidation of key physics tradeoffs, constraints, and uncertainties for the ultimate power-reactor embodiment

Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

Energy Technology Data Exchange (ETDEWEB)

Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-02-01

The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF₂) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR

Conceptual design of ICF reactor SENRI, Part II. Advances in design and pellet gain scaling

International Nuclear Information System (INIS)

Ido, S.; Mima, K.; Nakai, S.; Tsuji, R.; Yamanaka, C.

1984-01-01

This chapter reviews the recent design studies on reactor concepts with magnetically guided lithium flow, SENRI-I, SENRI-IA and SENRI-II. The routes from the present status to power reactors and an advanced fuel pellet concept is also discussed. Topics covered include pellet design, magnetohydrodynamic design of liquid lithium flow; reactor cavity concepts with magnetically guided lithium flow, a thermo-hydraulic analysis, a tritium recovery system; and an advanced fuel pellet concept for an inertial confinement fusion (ICF) reactor without a tritium breeding blanket. An advanced fuel pellet for an ICF reactor without a T breeder was studied in the model calculations, which showed sufficiently high values of pellet gain. Includes a table and 8 diagrams

Design study on sodium cooled large-scale reactor

International Nuclear Information System (INIS)

Murakami, Tsutomu; Hishida, Masahiko; Kisohara, Naoyuki

2004-07-01

In Phase 1 of the 'Feasibility Studies 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, design improvement for further cost reduction of establishment of the plant concept has been performed. This report summarizes the results of the design study on the sodium-cooled large-scale reactor performed in JFY2003, which is the third year of Phase 2. In the JFY2003 design study, critical subjects related to safety, structural integrity and thermal hydraulics which found in the last fiscal year has been examined and the plant concept has been modified. Furthermore, fundamental specifications of main systems and components have been set and economy has been evaluated. In addition, as the interim evaluation of the candidate concept of the FBR fuel cycle is to be conducted, cost effectiveness and achievability for the development goal were evaluated and the data of the three large-scale reactor candidate concepts were prepared. As a results 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 200,000 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)

The risks of nuclear energy technology. Safety concepts of light water reactors

Energy Technology Data Exchange (ETDEWEB)

Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Inst. fuer Kern- und Energietechnk (IKET); Kessler, Guenter; Veser, Anke; Schlueter, Franz-Hermann

2014-11-01

Analyses the risks of nuclear power stations. Discusses the security concept of reactors. Analyzes possible crash of air planes on a reactor containment. Presents measures against the spread of radioactivity after a severe accident. Written in engaging style for professionals and policy makers. The book analyses the risks of nuclear power stations. The security concept of reactors is explained. Measures against the spread of radioactivity after a severe accident, accidents of core melting and a possible crash of an air plane on a reactor containment are discussed. The book covers three scientific subjects of the safety concepts of Light Water Reactors: - A first part describes the basic safety design concepts of operating German Pressurized Water Reactors and Boiling Water Reactors including accident management measures introduced after the reactor accidents of Three Mile Island and Chernobyl. These safety concepts are also compared with the experiences of the Fukushima accidents. In addition, the safety design concepts of the future modern European Pressurized Water Reactor (EPR) and of the future modern Boiling Water Reactor SWR-1000 (KERENA) are presented. These are based on new safety research results of the past decades. - In a second, part the possible crash of military or heavy commercial air planes on a reactor containment is analyzed. It is shown that reactor containments can be designed to resist to such an airplane crash. - In a third part, an online decision system is presented. It allows to analyze the distribution of radioactivity in the atmosphere and to the environment after a severe reactor accident. It provides data for decisions to be taken by authorities for the minimization of radiobiological effects to the population. This book appeals to readers who have an interest in save living conditions and some understanding for physics or engineering.

Power reactor design trends

International Nuclear Information System (INIS)

Hogan, W.J.

1985-01-01

Cascade and Pulse Star represent new trends in ICF power reactor design that have emerged in the last few years. The most recent embodiments of these two concepts, and that of the HYLIFE design with which they will compare them, are shown. All three reactors depend upon protecting structural elements from neutrons, x rays and debris by injecting massive amounts of shielding material inside the reaction chamber. However, Cascade and Pulse Star introduce new ideas to improve the economics, safety, and environmental impact of ICF reactors. They also pose different development issues and thus represent technological alternatives to HYLIFE

Modular Stellarator Fusion Reactor concept

International Nuclear Information System (INIS)

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

1981-08-01

A preliminary conceptual study is made of the Modular Stellarator Reactor (MSR). A steady-state ignited, DT-fueled, magnetic fusion reactor is proposed for use as a central electric-power station. The MSR concept combines the physics of the classic stellarator confinement topology with an innovative, modular-coil design. Parametric tradeoff calculations are described, leading to the selection of an interim design point for a 4-GWt plant based on Alcator transport scaling and an average beta value of 0.04 in an l = 2 system with a plasma aspect ratio of 11. The physics basis of the design point is described together with supporting magnetics, coil-force, and stress computations. The approach and results presented herein will be modified in the course of ongoing work to form a firmer basis for a detailed conceptual design of the MSR

Molten salt reactor concept

International Nuclear Information System (INIS)

Sood, D.D.

1980-01-01

Molten salt reactor is an advanced breeder concept which is suited for the utilization of thorium for nuclear power production. This reactor is based on the use of solutions of uranium or plutonium fluorides in LiF-BeF 2 -ThF 4 as fuel. Unlike the conventional reactors, no external coolant is used in the reactor core and the fuel salt itself is circulated through heat exchangers to transfer the fission produced heat to a secondary salt (NaF-NaBF 4 ) for steam generation. A part of the fuel stream is continuously processed to isolate 233 Pa, so that it can decay to fissile 233 U without getting converted to 234 Pa, and for the removal of neutron absorbing fission products. This on-line processing scheme makes this reactor concept to achieve a breeding ratio of 1.07 which is the highest for any thermal breeder reactor. Experimental studies at the Bhabha Atomic Research Centre, Bombay, have established the use of plutonium as fuel for this reactor. This molten salt reactor concept is described and the work conducted at the Bhabha Atomic Research Centre is summarised. (auth.)

High Flux Isotope Reactor cold neutron source reference design concept

International Nuclear Information System (INIS)

Selby, D.L.; Lucas, A.T.; Hyman, C.R.

1998-05-01

In February 1995, Oak Ridge National Laboratory's (ORNL's) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH 2 ) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH 2 cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept

Evaluating usability of the Halden Reactor Large Screen Display. Is the Information Rich Design concept suitable for real-world installations?

International Nuclear Information System (INIS)

Braseth, Alf Ove

2013-01-01

Large Screen Displays (LSDs) are beginning to supplement desktop displays in modern control rooms, having the potential to display the big picture of complex processes. Information Rich Design (IRD) is a LSD concept used in many real-life installations in the petroleum domain, and more recently in nuclear research applications. The objectives of IRD are to provide the big picture, avoiding keyhole related problems while supporting fast visual perception of larger data sets. Two LSDs based on the IRD concept have been developed for large-scale nuclear simulators for research purposes; they have however suffered from unsatisfying user experience. The new Halden Reactor LSD, used to monitor a nuclear research reactor, was designed according to recent proposed Design Principles compiled in this paper to mitigate previously experienced problems. This paper evaluates the usability of the Halden Reactor LSD, comparing usability data with the replaced analogue panel, and data for an older IRD large screen display. The results suggest that the IRD concept is suitable for use in real-life applications from a user experience point of view, and that the recently proposed Design Principles have had a positive effect on usability. (author)

ULTRA SCWR+: Practical advanced water reactor concepts

International Nuclear Information System (INIS)

Duffey, Romney; Khartabil, Hussam; Kuran, Sermet; Zhou, Tracy; Pioro, Igor

2008-01-01

Modern thermal power plants now utilize supercritical steam cycles with thermal efficiencies of over 45%. Recent developments have lead to Ultra-SuperCritical (USC) systems, which adopt reheat turbines that can attain efficiencies of over 50%. Because these turbines are already developed, demonstrated and deployed worldwide, and use existing and traditional steam cycle technology, the simplest nuclear advance is to utilize these proven thermal cycle conditions by coupling this turbine type to a reactor. This development direction is fundamentally counter to the usual approach of adopting high-temperature gas-cooled (helium-cooled) reactor cycles, for which turbines have yet to be demonstrated on commercial scale unlike the supercritical steam turbines. The ULTRA (Ultra-supercritical Light water Thermal ReActor) SCWR+ concept adopts the fundamental design approach of matching a water and steam-cooled reactor to the ultra-supercritical steam cycle, adopting the existing and planned thermal power plant turbines. The HP and IP sections are fed with conditions of 25 MPa/625degC and 7 MPa/700degC, respectively, to achieve operating plant thermal efficiencies in excess of 50%, with a direct turbine cycle. By using such low-pressure reheated steam, this concept also adopts technology that was explored and used many years ago in existing water reactors, with the potential to produce large quantities of low cost heat, which can be used for other industrial and district processes. Pressure-Tube (PT) reactors are suitable for adoption of this design approach and, in addition, have other advantages that will significantly improve water-cooled reactor technology. These additional advantages include enhanced safety and improved resource utilization and proliferation resistance. This paper describes the PT-SCWR+ concept and its potential enhancements. (author)

The integral fast reactor concept

International Nuclear Information System (INIS)

Chang, Yoon I.; Marchaterre, J.F.

1987-01-01

The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) an integral fuel cycle, based on pyrometallurgical processing and injection-cast fuel fabrication, with the fuel cycle facility collocated with the reactor, if so desired. This paper gives a review of the IFR concept

Safety design study of fast breeder reactors in Japan

International Nuclear Information System (INIS)

Miura, M.; Inagaki, T.

1992-01-01

This paper reports on two fast breeder reactor (FBR) concepts, the tank type and the loop type, that have been studied as possible reactor designs to be used for a demonstration FBR (DFBR). The basic principle fo the DFBR design is to ensure plant safety through a defense-in-depth methodology. Improvements in the seismic and thermal stress designs have been attempted for both reactor concepts. The system design study strives to maximize the reliability of the safety-related systems and to rationalize commercialization of the plant

Design studies of Tokamak power reactor in JAERI

International Nuclear Information System (INIS)

Tone, T.; Nishikawa, M.; Tanaka, Y.

1985-01-01

Recent design studies of tokamak power reactor and related activities conducted in JAERI are presented. A design study of the SPTR (Swimming-Pool Type Reactor) concept was carried out in FY81 and FY82. The reactor design studies in the last two years focus on nuclear components, heat transport and energy conversion systems. In parallel of design studies, tokamak systems analysis code is under development to evaluate reactor performances, cost and net energy balance

Study of the reactor relevance of the NET design concept

International Nuclear Information System (INIS)

Reynolds, P.; Worraker, W.J.

1987-08-01

The objective of the study was to explore the reactor relevance of NET, i.e. whether the technologies and design principles proposed for NET can be directly extrapolated to a demonstration power reactor (DEMO). The main areas of study were those near to the plasma, namely the divertor, first wall and tritium breeding blanket. Other aspects which were investigated were tritium permeation and recovery, reactor maintenance, afterheat and effects of disruptions. The principal results of the study are briefly presented; the details of the work are given in fourteen appendices. These appendices were selected for INIS and indexed separately. The overall conclusion of the study is that the NET design is only partly relevant to the design requirements of a DEMO reactor. (U.K.)

Status of fusion reactor concept development in Japan

International Nuclear Information System (INIS)

Tsuji-Iio, Shunji

1996-01-01

Fusion power reactor studies in Japan based on magnetic confinement schemes are reviewed. As D-T fusion reactors, a steady-state tokamak reactor (SSTR) was proposed and extensively studied at the Japan Atomic Energy Research Institute (JAERI) and an inductively operated day-long tokamak reactor (IDLT) was proposed by a group at the University of Tokyo. The concept of a drastically easy maintenance (DREAM) tokamak reactor is being developed at JAERI. A high-field tokamak reactor with force-balanced coils as a volumetric neutron source is being studied by our group at Tokyo Institute of Technology. The conceptual design of a force-free helical reactor (FFHR) is under way at the National Institute for Fusion Science. A design study of a D- 3 He field-reversed configuration (FRC) fusion reactor called ARTEMIS was conducted by the FRC fusion working group of research committee of lunar base an lunar resources. (author)

Development project HTR-electricity-generating plant, concept design of an advanced high-temperature reactor steam cycle plant with spherical fuel elements (HTR-K)

International Nuclear Information System (INIS)

1978-07-01

The report gives a survey of the principal work which was necessary to define the design criteria, to determine the main design data, and to design the principal reactor components for a large steam cycle plant. It is the objective of the development project to establish a concept design of an edvanced steam cycle plant with a pebble bed reactor to permit a comparison with the direct-cycle-plant and to reach a decision on the concept of a future high-temperature nuclear power plant. It is tried to establish a largerly uniform basic concept of the nuclear heat-generating systems for the electricity-generating and the process heat plant. (orig.) [de

High Flux Isotope Reactor cold neutron source reference design concept

Energy Technology Data Exchange (ETDEWEB)

Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

1998-05-01

In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

Conceptual design of the SlimCS fusion DEMO reactor

International Nuclear Information System (INIS)

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Utoh, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; Sakurai, Shinji; Kurita, Genichi; Hayashi, Takao; Oyama, Naoyuki; Liu Changle; Hamamatsu, Kiyotaka; Inoue, Takashi; Ozeki, Takahisa; Sato, Masayasu; Suzuki, Satoshi; Kawashima, Hisato; Ezato, Koichiro; Tsuru, Daigo; Koizumi, Norikiyo; Sakamoto, Keiji; Ando, Masami; Sakamoto, Yoshiteru; Shibama, Yusuke; Suzuki, Takahiro; Takechi, Manabu; Takahashi, Koji; Hirose, Takanori; Sato, Satoru; Nozawa, Takashi; Tanigawa, Hisashi; Kakudate, Satoshi; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Ochiai, Kentaro; Ide, Shunsuke; Aiba, Nobuyuki; Shimizu, Katsuhiro; Honda, Mitsuru; Nakamichi, Masaru; Nishi, Hiroshi; Seki, Yoji; Nakamura, Yukiharu; Tsuchiya, Kunihiko; Yoshida, Tohru; Song Yuntao

2010-08-01

This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). Owing to low aspect ratio, the reactor will be capable of having comparatively high beta limit and high elongation (which can elevate the Greenwald density limit), having potential for high power density. The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m 2 . This report covers various aspects of design study including systematic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept. (author)

Planning a new research reactor for AECL: The MAPLE-MTR concept

International Nuclear Information System (INIS)

Lee, A.G.; Lidstone, R.F.; Donnelly, J.V.

1992-01-01

AECL Research is assessing its needs and options for future irradiation research facilities. A planning team has been assembled to identify the irradiation requirements for AECL's research programs and compile options for satisfying the irradiation requirements. The planning team is formulating a set of criteria to evaluate the options and will recommend a plan for developing an appropriate research facility. Developing the MAPLE Materials Test Reactor (MAPLE-MTR) concept to satisfy AECL's irradiation requirements is one option under consideration by the planning team. AECL is undertaking this planning phase because the NRU reactor is 35 years old and many components are nearing the end of their design life. This reactor has been a versatile facility for proof testing CANDU components and fuel designs because the CANDU irradiation environment was simulated quite well. However, the CANDU design has matured and the irradiation requirements have changed. Future research programs will emphasize testing CANDU components near or beyond their design limits. To provide these irradiation conditions, the NRU reactor needs to be upgraded. Upgrading and refurbishing the NRU reactor is being considered, but the potentially large costs and regulatory uncertainties make this option very challenging. AECL is also developing the MAPLE-MTR concept as a potential replacement for the NRU reactor. The MAPLE-MTR concept starts from the recent MAPLE-X10 design and licensing experience and adapts this technology to satisfy the primary irradiation requirements of AECL's research programs. This approach should enable AECL to minimize the need for major advances in nuclear technology (e.g., fuel design, heat transfer). The preliminary considerations for developing the MAPLE-MTR concept are presented in this report. A summary of AECL's research programs is presented along with their irradiation requirements. This is followed by a description of safety criteria that need to be taken into

Basic conceptions for reactor pressure vessel manipulators and their evaluation

International Nuclear Information System (INIS)

Popp, P.

1987-01-01

The study deals with application fields and basic design conceptions of manipulators in reactor pressure vessels as well as their evaluation. It is shown that manipulators supported at the reactor flange have essential advantages

Systemization of Design and Analysis Technology for Advanced Reactor

International Nuclear Information System (INIS)

Kim, Keung Koo; Lee, J.; Zee, S. K.

2009-01-01

The present study is performed to establish the base for the license application of the original technology by systemization and enhancement of the technology that is indispensable for the design and analysis of the advanced reactors including integral reactors. Technical reports and topical reports are prepared for this purpose on some important design/analysis methodology; design and analysis computer programs, structural integrity evaluation of main components and structures, digital I and C systems and man-machine interface design. PPS design concept is complemented reflecting typical safety analysis results. And test plans and requirements are developed for the verification of the advanced reactor technology. Moreover, studies are performed to draw up plans to apply to current or advanced power reactors the original technologies or base technologies such as patents, computer programs, test results, design concepts of the systems and components of the advanced reactors. Finally, pending issues are studied of the advanced reactors to improve the economics and technology realization

Design study on evaluation for power conversion system concepts in high temperature gas cooled reactor with gas turbine

International Nuclear Information System (INIS)

Minatsuki, Isao; Mizokami, Yorikata

2007-01-01

The design studies on High Temperature Gas Cooled Reactor with Gas Turbine (HTGR-GT) have been performed, which were mainly promoted by Japan Atomic Energy Agency (JAEA) and supported by fabricators in Japan. HTGR-GT plant feature is almost determined by selection of power conversion system concepts. Therefore, plant design philosophy is observed characteristically in selection of them. This paper describes the evaluation and analysis of the essential concepts of the HTGR-GT power conversion system through the investigations based on our experiences and engineering knowledge as a fabricator. As a result, the following concepts were evaluated that have advantages against other competitive one, such as the horizontal turbo machine rotor, the turbo machine in an individual vessel, the turbo machine with single shaft, the generator inside the power conversion vessel, and the power conversion system cycle with an intercooler. The results of the study can contribute as reference data when the concepts will be selected. Furthermore, we addressed reasonableness about the concept selection of the Gas Turbine High Temperature Reactor GTHTR300 power conversion system, which has been promoted by JAEA. As a conclusion, we recognized the GTHTR300 would be one of the most promising concepts for commercialization in near future. (author)

Extrap conceptual fusion reactor design study

International Nuclear Information System (INIS)

Eninger, J.E; Lehnert, B.

1987-12-01

A study has recently been initiated to asses the fusion reactor potential of the Extrap concept. A reactor model is defined that fulfills certain economic and environmental criteria. This model is applied to Extrap and a reference reactor is outlined. The design is optimized by varying parameters subject to both physics and engineering constraints. Several design options are examined and key engineering issues are identified and addressed. Some preliminary results and conclusions of this work are summarized. (authors)

A new advanced safe nuclear reactor concept

International Nuclear Information System (INIS)

Sefidvash, Farhang

1999-01-01

The reactor design is based on fluidized bed concept and utilizes pressurized water reactor technology. The fuel is automatically removed from the reactor by gravity under any accident condition. The reactor demonstrates the characteristics of inherent safety and passive cooling. Here two options for modification to the original design are proposed in order to increase the stability and thermal efficiency of the reactor. A modified version of the reactor involves the choice of supercritical steam as the coolant to produce a plant thermal efficiency of about 40%. Another is to modify the shape of the reactor core to produce a non-fluctuating bed and consequently guarantee the dynamic stability of the reactor. The mixing of Tantalum in the fuel is also proposed as an additional inhibition to power excursion. The spent fuel pellets may not be considered nuclear waste since they are in the shape and size that can easily be used as a a radioactive source for food irradiation and industrial applications. The reactor can easily operate with any desired spectrum by varying the porosity in order to be a plutonium burner or utilize a thorium fuel cycle. (author)

A multi-crucible core-catcher concept: Design considerations and basic results

International Nuclear Information System (INIS)

Szabo, I.

1995-01-01

A multi-crucible core-catcher concept to be implemented in new light water reactor containments has recently been proposed. This paper deals with conceptual design considerations and the various ways this type of core-catcher could be designed to meet requirements for reactor application. A systematic functional analysis of the multi-crucible core-catcher concept and the results of the preliminary design calculation are presented. Finally, the adequacy of the multi-crucible core-catcher concept for reactor application is discussed. (orig.)

Preliminary Study for Conceptual Design of Advanced Long Life Small Modular Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Kim, T. K. [Argonne National Laboratory, Argonne (United States)

2015-05-15

As one of the non-water coolant Small-Modular Reactor (SMR) core concepts for use in the mid- to long-term, ANL has proposed a 100 MWe Advanced sodium-cooled Fast Reactor core concept (AFR-100) targeting a small grid, transportable from pre-licensed factories to the remote plant site for affordable supply. Various breed-and-burn core concepts have been proposed to extend the reactor cycle length, which includes CANDLE with a cigar-type depletion strategy, TerraPower reactors with fuel shuffling for effective breeding, et al. UNIST has also proposed an ultra-long cycle fast reactor (UCFR) core concept having the power rating of 1000 MWe. By adopting the breed-and-burn strategies, the UCFR core can maintain criticality for a targeting reactor lifetime of 60 years without refueling. The objective of this project is to develop an advanced long-life SMR core concept by adopting both the small modular design features of the AFR-100 and the long-life breed-and-burn concept of the UCFR. A conceptual design of long life small modular fast reactor is under development by adopting both the small modular design features of the AFR-100 and the long-life breed-and-burn concept of the UCFR. The feasibility of the long-life fast reactor concepts was reviewed to obtain the core design guidelines and the reactor design requirements of long life small modular fast reactor were proposed in this study.

Particle bed reactor nuclear rocket concept

International Nuclear Information System (INIS)

Ludewig, H.

1991-01-01

The particle bed reactor nuclear rocket concept consists of fuel particles (in this case (U,Zr)C with an outer coat of zirconium carbide). These particles are packed in an annular bed surrounded by two frits (porous tubes) forming a fuel element; the outer one being a cold frit, the inner one being a hot frit. The fuel element are cooled by hydrogen passing in through the moderator. These elements are assembled in a reactor assembly in a hexagonal pattern. The reactor can be either reflected or not, depending on the design, and either 19 or 37 elements, are used. Propellant enters in the top, passes through the moderator fuel element and out through the nozzle. Beryllium used for the moderator in this particular design to withstand the high radiation exposure implied by the long run times

Fission fragment assisted reactor concept for space propulsion: Foil reactor

International Nuclear Information System (INIS)

Wright, S.A.

1991-01-01

The concept is to fabricate a reactor using thin films or foils of uranium, uranium oxide and then to coat them on substrates. These coatings would be made so thin as to allow the escaping fission fragments to directly heat a hydrogen propellant. The idea was studied of direct gas heating and direct gas pumping in a nuclear pumped laser program. Fission fragments were used to pump lasers. In this concept two substrates are placed opposite each other. The internal faces are coated with thin foil of uranium oxide. A few of the advantages of this technology are listed. In general, however, it is felt that if one look at all solid core nuclear thermal rockets or nuclear thermal propulsion methods, one is going to find that they all pretty much look the same. It is felt that this reactor has higher potential reliability. It has low structural operating temperatures, very short burn times, with graceful failure modes, and it has reduced potential for energetic accidents. Going to a design like this would take the NTP community part way to some of the very advanced engine designs, such as the gas core reactor, but with reduced risk because of the much lower temperatures

Supercritical-pressure, once-through cycle light water cooled reactor concept

International Nuclear Information System (INIS)

Oka, Yoshiaki; Koshizuka, Seiichi

2001-01-01

The purpose of the study is to develop new reactor concepts for the innovation of light water reactors (LWR) and fast reactors. Concept of the once-through coolant cycle, supercritical-pressure light water cooled reactor was developed. Major aspects of reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the reactor is monitoring coolant flow rate instead of water level of LWR. The reactor system is compact and simple because of high specific enthalpy of supercritical water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the high outlet coolant temperature. The reactor is compatible with tight fuel lattice fast reactor because of the high head pumps and low coolant flow rate. The power rating of the fast reactor is higher than the that of thermal reactor because of the high power density. (author)

Mechanical design of a magnetic fusion production reactor

International Nuclear Information System (INIS)

Neef, W.S.; Jassby, D.L.

1986-01-01

The mechanical aspects of a tandem mirror and tokamak concepts for the tritium production mission are compared, and a proposed breeding blanket configuration for each type of reactor is presented in detail, along with a design outline of the complete fusion reaction system. In both cases, the reactor design is developed sufficiently to permit preliminary cost estimates of all components. A qualitative comparison is drawn between both concepts from the view of mechanical design and serviceability, and suggestions are made for technology proof tests on unique mechanical features. Detailed cost breakdowns indicate less than 10% difference in the overall costs of the two reactors

Accident analysis for new reactor concepts and VVER type reactor design with advanced fuel. STC with Russia. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.; Mittag, S.; Rohde, U.; Seidel, A.

2000-10-01

In the frame of a project on scientific-technical cooperation funded by BMBF/BMWi, the 3D reactor dynamics code DYN3D developed at Forschungszentrum Rossendorf (FZR), has been transferred to the Institute of Physics and Power Engineering (IPPE) Obninsk in Russia and integrated into the software package of IPPE. DYN3D has been coupled to a thermohydraulic system code used in IPPE making available 3D neutron kinetics within this software package. A new macroscopic cross section library has been created using a modified version of the WIMS/D4 code. This library includes data for modernized fuel design containing burnable absorbers in different concentrations, which is tested in VVER-1000 type reactors. The cross section library has been connected to DYN3D. Calculations were performed to check the library in comparison with other data libraries and codes. The code DYN3D and the coupled 3D neutron kinetics/thermal hydraulics code system were used to perform analyses of Anticipated Transients Without Scram (ATWS) for the reactor design ABV-67, an integral reactor concept with small power developed under participation of IPPE. The fluid dynamics code DINCOR developed at IPPE was transferred to FZR. It was used in validation calculations on test problems for the short-term core melt behaviour (CORVIS experiments). (orig.) [de

ELMO Bumpy Torus fusion-reactor design study

International Nuclear Information System (INIS)

Bathke, C.G.; Krakowski, R.A.

1981-01-01

A complete power plant design of a 1200-MWe ELMO Bumpy Torus Reactor (EBTR) is described that emphasizes those features that are unique to the EBT confinement concept, with subsystems and balance-of-plant items that are generic to magnetic fusion being adopted from past, more extensive tokamak reactor designs

Application of the integrated blanket-coil concept (IBC) to fusion reactors

International Nuclear Information System (INIS)

Embrechts, M.J.; Steiner, D.; Mohanti, R.; Duggan, W.

1987-01-01

A novel concept is proposed for combining the blanket and coil functions of a fusion reactor into a single component and several unique applications to fusion reactor embodiments are identified. The proposed concept takes advantage of the fact that lithium is a good electrical conductor in addition to being a unique tritium-breeding material capable of energy recovery and transport at high temperatures. This concept, designated the ''integrated-blanket-coil (IBC) concept'' has the potential for: allowing fusion reactor embodiments which are easier to maintain; making fusion reactors more compact with an intrinsic ultra-high mass power density (net kW/sub E//metric tonne); and enhancing the tritium breeding potential for special coil applications such as ohmic heating and bean identation. By assuming a sandwich construction for the IBC walls (i.e., a layered combination of a thin wall of structural material, insulator and structural materials) the magnetohydrodynamic (MHD)-induced pressure drops and associated pressure stresses are modest and well below design limits. Possible unique applications of the IBC concept have been investigated and include the IBC concept applied to the poloidal field (PF) coils, toroidal field (TF) coils, divertor coils, ohmic heating (OH) coils, and identation coils for bean shaping

International tokamak reactor conceptual design overview

International Nuclear Information System (INIS)

Stacey, W.M. Jr.

1983-01-01

The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and the USSR, under the auspices of the IAEA, to assess, define, design, construct and operate the next major experiment in the World Tokamak Program beyond the TFTR, JET, JT-60, T-15 generation. During the Zero-Phase (1979), a technical data base assessment was performed, leading to a positive assessment of feasibility. During Phase-I (1/80-6/81), a conceptual design was developed to define the concept. The programmatic objectives are that INTOR should: (1) be the maximum reasonable step beyond the TFTR, JET, JT-60, T-15 generation of tokamaks, (2) demonstrate the plasma performance required for tokamak DEMOs, (3) test the development and integration into a reactor system of those technologies required for a DEMO, (4) serve as a test facility for blanket, tritium production, materials, and plasma engineering technology, (5) test fusion reactor component reliability, (6) test the maintainability of a fusion reactor, and (7) test the factors affecting the reliability, safety and environmental acceptability of a fusion reactor. A conceptual design has been developed to define a device which is consistent with these objectives. The design concept could, with a reasonable degree of confidence, be developed into a workable engineering design of a tokamak that met the performance objectives of INTOR. There is some margin in the design to allow for uncertainty. While design solutions have been found for all of the critical issues, the overall design may not yet be optimal. (author)

Design of a new research reactor : 1st year conceptual design

International Nuclear Information System (INIS)

Park, Cheol; Lee, B. C.; Chae, H. T.

2004-01-01

A new research reactor model satisfying the strengthened regulatory environments and the changed circumstances around nuclear society should be prepared for the domestic and international demand of research reactor. This can also lead to the improvement of technologies and fostering manpower obtained during the construction and the operation of HANARO. In this aspect, this study has been launched and the 1st year conceptual design has been carried out in 2003. The major tasks performed at the first year of conceptual design stage are as follows; Establishments of general design requirements of research reactors and experimental facilities, Establishment of fuel and reactor core concepts, Preliminary analysis of reactor physics and thermal-hydraulics for conceptual core, Conceptual design of reactor structure and major systems, International cooperation to establish foundations for exporting

Evaluation of the breed/burn fast reactor concept

International Nuclear Information System (INIS)

Atefi, B.; Driscoll, M.J.; Lanning, D.D.

1979-12-01

A core design concept and fuel management strategy, designated breed/burn, has been evaluated for heterogeneous fast breeder reactors. In this concept internal blanket assemblies after fissile material is bred in over several incore cycles, are shuffled into a moderated radial blanket and/or central island. The most promising materials combination identified used thorium in the internal blankets (due to the superior performance of epithermal Th-U233 systems) and zirconium hydride (ZrH 16 ) as the moderator

Containment concepts assessment for the SEAFP reactor

International Nuclear Information System (INIS)

Di Pace, L.; Natalizio, A.

2000-01-01

A simple methodology has been developed for making relative comparisons of potential containment designs for future fusion reactors. The assessment methodology requires only conceptual design information. The application of this methodology, at the early stages of a fusion reactor design, provides designers useful information regarding the suitability of various containment designs and design features. Because the radiation hazard from the operation of future fusion power reactors is expected to be low, the containment design, in addition to public safety, needs to take into account worker safety considerations, as well as factors important to the reliable and economical operation of the power plant. Several containment concepts have been assessed with a methodology that takes into account public safety, worker safety, operability and maintainability as well as cost. This paper describes this methodology and presents the results of the assessment. The paper concludes that, to obtain a containment design that is optimised with respect to safety, operational and cost factors, designers should focus on a containment that is conceptually simple-that is, one utilising a single, large containment building without relying on special features such as expansion volumes, pressure suppression pools or spray systems

Innovative reactor core: potentialities and design

International Nuclear Information System (INIS)

Artioli, C.; Petrovich, Carlo; Grasso, Giacomo

2010-01-01

Gen IV nuclear reactors are considered a very attractive answer for the demand of energy. Because public acceptance they have to fulfil very clearly the requirement of sustainable development. In this sense a reactor concept, having by itself a rather no significant interaction with the environment both on the front and back end ('adiabatic concept'), is vital. This goal in mind, a new way of designing such a core has to be assumed. The starting point must be the 'zero impact'. Therefore the core will be designed having as basic constraints: a) fed with only natural or depleted Uranium, and b) discharges only fission products. Meantime its potentiality as a net burner of Minor Actinide has to be carefully estimated. This activity, referred to the ELSY reactor, shows how to design such an 'adiabatic' core and states its reasonable capability of burning MA legacy in the order of 25-50 kg/GW e y. (authors)

Fourth Generation Reactor Concepts

International Nuclear Information System (INIS)

Furtek, A.

2008-01-01

Concerns over energy resources availability, climate changes and energy supply security suggest an important role for nuclear energy in future energy supplies. So far nuclear energy evolved through three generations and is still evolving into new generation that is now being extensively studied. Nuclear Power Plants are producing 16% of the world's electricity. Today the world is moving towards hydrogen economy. Nuclear technologies can provide energy to dissociate water into oxygen and hydrogen and to production of synthetic fuel from coal gasification. The introduction of breeder reactors would turn nuclear energy from depletable energy supply into an unlimited supply. From the early beginnings of nuclear energy in the 1940s to the present, three generations of nuclear power reactors have been developed: First generation reactors: introduced during the period 1950-1970. Second generation: includes commercial power reactors built during 1970-1990 (PWR, BWR, Candu, Russian RBMK and VVER). Third generation: started being deployed in the 1990s and is composed of Advanced LWR (ALWR), Advanced BWR (ABWR) and Passive AP600 to be deployed in 2010-2030. Future advances of the nuclear technology designs can broaden opportunities for use of nuclear energy. The fourth generation reactors are expected to be deployed by 2030 in time to replace ageing reactors built in the 1970s and 1980s. The new reactors are to be designed with a view of the following objectives: economic competitiveness, enhanced safety, minimal radioactive waste production, proliferation resistance. The Generation IV International Forum (GIF) was established in January 2000 to investigate innovative nuclear energy system concepts. GIF members include Argentina, Brazil, Canada, Euratom, France Japan, South Africa, South Korea, Switzerland, United Kingdom and United States with the IAEA and OECD's NEA as permanent observers. China and Russia are expected to join the GIF initiative. The following six systems

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Evaluation of national and foreign design and application concepts for small-sized high-temperature reactors

International Nuclear Information System (INIS)

Hahn, L.; Nockenberg, B.

1990-03-01

There are currently the design types Modular HTR, MHTGR, HTTR, HTR-500, and GHR. The different design concepts are explained and are analysed with regard to their application potentials in Germany and abroad, their export into developing countries, their general and specific safety aspects, the relevant fuel cycle aspects (fuel supply and spent fuel management), and the resulting problems arising in connection with the principle of non-proliferation. In the concluding section of the report, misgivings are stated as to possible military applications of this reactor type. (DG) [de

Status of fusion reactor blanket design

International Nuclear Information System (INIS)

Smith, D.L.; Sze, D.K.

1986-02-01

The recent Blanket Comparison and Selection Study (BCSS), which was a comprehensive evaluation of fusion reactor blanket design and the status of blanket technology, serves as an excellent basis for further development of blanket technology. This study provided an evaluation of over 130 blanket concepts for the reference case of electric power producing, DT fueled reactors in both Tokamak and Tandem Mirror (TMR) configurations. Based on a specific set of reactor operating parameters, the current understanding of materials and blanket technology, and a uniform evaluation methodology developed as part of the study, a limited number of concepts were identified that offer the greatest potential for making fusion an attractive energy source

Design concepts and safety concerns of the small and medium size reactors (SMR)

International Nuclear Information System (INIS)

Seul, Kwang Won; Lee, Jae Hun; Kim, Hho Jung

1998-01-01

The small and medium size reactors (SMR) and interface facilities such as desalination plant are expected to be located near the population area because of restrictions in transporting the plant products such as fresh water to long distance area. To protect the public around the plant facility from the possible release of radioactive materials, the design development of the SMR is focusing on an enhancement of the safety and reliability as well as the economics. In this study, the major safety concepts of the SMR designs significantly different from the current PWR designs are investigated and the safety concerns applicable to the integrated SMR design of Korea (called SMART), were identified. Those safety issues include the use of proven technology, application of strengthening defense in depth, event categorization and selection, simplification of emergency planning, determination of accident source terms and so on. The efforts to resolve the safety concerns in the design stage will provide an improvement of the safety of the SMART design

Reactor concepts for laser fusion

International Nuclear Information System (INIS)

Meier, W.R.; Maniscalco, J.A.

1977-07-01

Scoping studies were initiated to identify attractive reactor concepts for producing electric power with laser fusion. Several exploratory reactor concepts were developed and are being subjected to our criteria for comparing long-range sources of electrical energy: abundance, social costs, technical feasibility, and economic competitiveness. The exploratory concepts include: a liquid-lithium-cooled stainless steel manifold, a gas-cooled graphite manifold, and fluidized wall concepts, such as a liquid lithium ''waterfall'', and a ceramic-lithium pellet ''waterfall''. Two of the major reactor vessel problems affecting the technical feasibility of a laser fusion power plant are: the effects of high-energy neutrons and cyclical stresses on the blanket structure and the effects of x-rays and debris from the fusion microexplosion on the first-wall. The liquid lithium ''waterfall'' concept is presented here in more detail as an approach which effectively deals with these damaging effects

Development of core thermal-hydraulics module for intelligent reactor design system (IRDS)

International Nuclear Information System (INIS)

Kugo, Teruhiko; Nakagawa, Masayuki; Fujii, Sadao.

1994-08-01

We have developed an innovative reactor core thermal-hydraulics module where a designer can easily and efficiently evaluate his design concept of a new type reactor in the thermal-hydraulics field. The main purpose of this module is to decide a feasible range of basic design parameters of a reactor core in a conceptual design stage of a new type reactor. The module is to be implemented in Intelligent Reactor Design System (IRDS). The module has the following characteristics; 1) to deal with several reactor types, 2) four thermal hydraulics and fuel behavior analysis codes are installed to treat different type of reactors and design detail, 3) to follow flexibly modification of a reactor concept, 4) to provide analysis results in an understandable way so that a designer can easily evaluate feasibility of his concept, and so on. The module runs on an engineering workstation (EWS) and has a user-friendly man-machine interface on a pre- and post-processing. And it is equipped with a function to search a feasible range called as Design Window, for two design parameters by artificial intelligence (AI) technique and knowledge engineering. In this report, structure, guidance for users of an usage of the module and instruction of input data for analysis modules are presented. (author)

Design of megawatt power level heat pipe reactors

Energy Technology Data Exchange (ETDEWEB)

Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dasari, Venkateswara Rao [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reid, Robert Stowers [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-12

An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors. The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.

Reactor design concepts for radiation processing

International Nuclear Information System (INIS)

Berejka, A.J.

2004-01-01

During the formative years of irradiation processing, the 1950s and 1960s, there was laboratory and academic interest in the use of this form of energy transfer to initiate polymerization for the manufacture of plastics and in other chemical processes. Studies were often based on low-dose-rate Cobalt-60 systems. The electron beam (EB) accelerator technology of the time was not as yet at the robust and industrially reliable state that it is now at the beginning of the twenty-first century. A series of reactor designs illustrate how an electron beam can be incorporated into reactor vessels for initiating gas and liquid phase polymerizations on a continuous basis. Development of such approaches, which would rely upon contemporary, high current electron beams to initiate polymerization, would help the chemical processing industry alleviate its problems of catalyst disposal and its related environmental concerns. Systems for treating materials in bulk at low doses, such as those typically used for grain disinfection, at high through-put rates, are also illustrated. Simplified shielding is envisioned in each proposed process system

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

Study on concept of web-based reactor piping design data platform

International Nuclear Information System (INIS)

Wang Yu; Zhou Yu; Dong Jianling; Meng Yang

2005-01-01

For solving the piping design problems such as design data deficiency, designer communication inconvenience and design project inconsistence, Reactor Piping Design Database Platform, which is the main part of the Integrated Nuclear Project Research Platform, is proposed by analyzing the nuclear piping designs in detail. The functions and system structures of the platform are described in the paper for the sake of the realization of the Reactor Piping Design Database Platform. The platform is constituted by web-based management interface, AutoPlant selected as CAD software, and relation database management system (DBMS). (authors)

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

International Nuclear Information System (INIS)

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

2012-01-01

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

Conceptual design of multipurpose compact research reactor

International Nuclear Information System (INIS)

Nagata, Hiroshi; Kusunoki, Tsuyoshi; Hori, Naohiko; Kaminaga, Masanori

2012-01-01

Conceptual design of the high-performance and low-cost multipurpose compact research reactor which will be expected to construct in the nuclear power plant introduction countries, started from 2010 in JAEA and nuclear-related companies in Japan. The aims of this conceptual design are to achieve highly safe reactor, economical design, high availability factor and advanced irradiation utilization. One of the basic reactor concept was determined as swimming pool type, thermal power of 10MW and water cooled and moderated reactor with plate type fuel element same as the JMTR. It is expected that the research reactors are used for human resource development, progress of the science and technology, expansion of industry use, lifetime extension of LWRs and so on. (author)

Conceptual design study of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1986-11-01

Since 1980 the design study has been conducted at JAERI for the Fusion Experimental Reactor (FER) which has been proposed to be the next machine to JT-60 in the Japanese long term program of fusion reactor development. During two years from 1984 to 1985 FER concept was reviewed and redesigned. This report is the summary of the results obtained in the review and redesign activities in 1984 and 85. In the first year FER concept was discussed again and its frame work was reestablished. According to the new frame work the major reactor components of FER were designed. In the second year the whole plant system design including plant layout plan was conducted as well as the more detailed design analysis of the reactor conponents. The newly established frame for FER design is as follows: 1) Plasma : Self-ignition. 2) Operation scenario : Quasi-steady state operation with long burn pulse. 3) Neutron fluence on the first wall : 0.3 MWY/M 2 . 4) Blanket : Non-tritium breeding blanket with test modules for breeding blanket development. 5) Magnets : Superconducting Magnets. (author)

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)

Identification of improvements of advanced light water reactor concepts

International Nuclear Information System (INIS)

Frisch, W.; Liesch, K.; Riegel, B.

1993-01-01

The scope of this report is to identify the improvement of reactor developments with respect to reactor safety. This includes the collection of non-proprietary information on the description of the advanced design characteristics, especially summary design descriptions and general publications. This documentation is not intended to include a safety evaluation of the advanced concepts; however, it is structured in such a way that it can serve as a basis for a future safety evaluation. This is taken into account in the structure of the information regarding the distinction of the various concepts with respect to their 'advancement' and the classification of design characteristics according to some basic safety aspects. The overall description concentrates on those features which are relevant to safety. Other aspects, such as economy, operational features, maintenance, the construction period, etc...are not considered explicitly in this report

Modified-open fuel cycle performance with breed-and-burn advanced reactor concepts

International Nuclear Information System (INIS)

Heidet, Florent; Kim, Taek K.; Taiwo, Temitope A.

2011-01-01

Recent advances in fast reactor designs enable significant increase in the uranium utilization in an advanced fuel cycle. The category of fast reactors, collectively termed breed-and-burn reactor concepts, can use a large amount of depleted uranium as fuel without requiring enrichment with the exception of the initial core critical loading. Among those advanced concepts, some are foreseen to operate within a once-through fuel cycle such as the Traveling Wave Reactor, CANDLE reactor or Ultra-Long Life Fast Reactor, while others are intended to operate within a modified-open fuel cycle, such as the Breed-and-Burn reactor and the Energy Multiplier Module. This study assesses and compares the performance of the latter category of breed-and-burn reactors at equilibrium state. It is found that the two reactor concepts operating within a modified-open fuel cycle can significantly improve the sustainability and security of the nuclear fuel cycle by decreasing the uranium resources and enrichment requirements even further than the breed-and-burn core concepts operating within the once-through fuel cycle. Their waste characteristics per unit of energy are also found to be favorable, compared to that of currently operating PWRs. However, a number of feasibility issues need to be addressed in order to enable deployment of these breed-and-burn reactor concepts. (author)

Design approaches for enhancing the engineering feasibility of tokamak power reactors

International Nuclear Information System (INIS)

Shannon, T.E.; Steiner, D.

1977-01-01

The design approach developed in the ORNL Fusion Power Demonstration Study is reviewed. The design concepts having greatest impact on reactor feasibility by the application of current or near term technology are described briefly. These are: blanket structural material, blanket coolant, power conversion system, and pulsed electrical system. Concepts relative to the approach taken to simplify the overall reactor design are listed

Reactor design for nuclear electric propulsion

International Nuclear Information System (INIS)

Koenig, D.R.; Ranken, W.A.

1979-01-01

Conceptual design studies of a nuclear power plant for electric propulsion of spacecrafts have been on going for several years. An attractive concept which has evolved from these studies and which has been described in previous publications, is a heat-pipe cooled, fast spectrum nuclear reactor that provides 3 MW of thermal energy to out-of-core thermionic converters. The primary motivation for using heat pipes is to provide redundancy in the core cooling system that is not available in gas or liquid-metal cooled reactors. Detailed investigation of the consequences of heat pipe failures has resulted in modifications to the basic reactor design and has led to consideration of an entirely different core design. The new design features an integral laminated core configuration consisting of alternating layers of UO 2 and molybdenum sheets that span the entire diameter of the core. Design characteristics are presented and compared for the two reactors

International standardization of nuclear reactor designs - the way forward

International Nuclear Information System (INIS)

Raetzke, Christian

2010-01-01

The concept of 'International Standardization of Nuclear Reactor Designs' means that vendors could build their designs in every country without having to adapt it specifically to national safety requirements. Such standardization would have two main effects. It would greatly facilitate nuclear new build worldwide by giving greater efficiency and certainty to the national licensing procedures; by taking into account the fact that vendors, and nowadays also utilities, are active across borders; by helping developing countries to establish their nuclear new build programmes; and by reducing the strain on human resources on both the regulators' and the industry's side. The second valuable effect of standardization would be to further enhance safety by improving the exchange of construction and operating experience among a number of reactors belonging to fleets of the same design. The World Nuclear Association's CORDEL (Cooperation in Reactor Design Evaluation and Licensing) Group has developed a concept for implementation of international standardization of reactor designs. It has defined a number of steps to be taken by industry. At the same time, possibilities offered by national and international regulatory mechanisms would have to be fully made use of, and some changes in regulatory frameworks might be necessary. Some steps especially towards greater cooperation of regulators have already been taken; however, much still remains to be done. The concept of deploying standardized reactor designs across a number of countries supposes an alignment and, if possible, harmonization of national safety standards; a streamlining of national licensing procedures, making them more efficient and predictable; and the willingness of national regulators to take into account licensing done in other countries. In the end, this should lead to a mutual acceptance of design approvals or, in a more distant future, even to a multinational design approval process. All in all, the concept

Establishment of design concept of large capacity passive reactor KP1000 and performance evaluation of safety system for LBLOCA

Energy Technology Data Exchange (ETDEWEB)

Kim, Seong O.; Hwang, Young Dong; Kim, Young In; Chang, Moon Hee

1997-03-01

This study was performed to establish the design concepts and to evaluate the performance of safety features of large capacity passive reactor (1000 MWe grade). The design concepts of the large capacity passive reactor `KP1000` were established to generate 1000 MW electric power based on the AP600 of Westinghouse by increasing the number of reactor coolant loop and by increasing the size of reactor internals/core. To implement the analysis of the LBLOCA for KP1000, various kinds of computer codes being considered, it was concluded that RELAP5 was the most appropriate one in availability and operations in present situation. By the analysis of the computer code `RELAP5/Mod3.2.1.2`, following conclusions were derived as described below. First, by spectrum analysis of the discharge factor of the berak part, the most conservative discharge factor C{sub D}=1.2 and the PCT value of KP1000 was 1254F, which is slightly higher than the value of AP600 but is much less than the existing active reactor `Kori 3 and 4` where blowdown PCT value is 1693.4 deg F and reflooding PCT is 1918.4 deg F. Second, after the 200 seconds from the initiation of LBLOCA, IRWST water was supplied in a stable state and the maximum temperature of clad were maintained in a saturated condition. Therefore, it was concluded that the passive safety features of KP1000 keep reactor core from being damaged for large break LOCA. (author). 11 refs., 28 tabs., 37 figs.

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-01-01

Conceptual Design of Fusion Experimental Reactor (FER) of which the objective will be to realize self-ignition with D-T reaction is reported. Mechanical Configurations of FER are characterized with a noncircular plasma and a double-null divertor. The primary aim of design studies is to demonstrate fissibility of reactor structures as compact and simple as possible with removable torus sectors. The structures of each component such as a first-wall, blanket, shielding, divertor, magnet and so on have been designed. It is also discussed about essential reactor plant system requirements. In addition to the above, a brief concept of a steady-state reactor based on RF current drive is also discussed. The main aim, in this time, is to examine physical studies of a possible RF steady-state reactor. (author)

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

Comparative analysis of sub-critical transmutation reactor concepts

International Nuclear Information System (INIS)

Chang, S. H.

1997-01-01

The long-lived nuclear wastes have been substantially generated from the light water reactor for a few decades. The toxicity of these spent fuels will be higher than that of the uranium ore, even if those will be stored in the repository more than ten thousands. Hence the means of transmuting the key long-lived nuclear wastes, primarily the minor actinides, using a hybrid proton accelerator and subcritical transmutation reactor, are proposed. Until now, the representative concepts for a subcritical transmutation reactor are the Energy Amplifier, the OMEGA project, the ATW and the MSBR. The detailed concepts and the specifications are illustrated in Table 1. The design requirements for the subcritical transmutation reactor are the high transmutation rate of long-lived nuclear wastes, safety and economics. And to propose the subcritical transmutation reactor concepts, the coolant, the target material and fuel type are carefully considered. In these aspects, the representative concepts for a subcritical transmutation reactor in Table 1 have been surveyed. The requirements for a target and a coolant are the reliable, low maintenance operation and safe operation to minimize the wastes. The reliable, low maintenance operation and safe operation to minimize the wastes. The reliable coolant must have the low melting point, high heat capacity and excellent physical properties. And the target material must have high neutron yield for a given proton condition and easy heat removal capability. Therefore in respect with the above requirements, Pb-Bi is proposed as the coolant and the target material for the subcritical reactor. Because the neutron yield for a given proton energy increases linearly with mass number up to bismuth but in heavier elements spallation events sharply increase both the neutron and heat outputs, Pb-Bi meets not only such the requirements as the above for the coolant but also those for the coolant and target, the simplification of system can be achieved

High-Temperature Gas-Cooled Test Reactor Point Design

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-01

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

Development of design technology for advanced pressurized water reactor

International Nuclear Information System (INIS)

Kim, Si Hwan; Chang, Moon Hee; Lee, Jong Chul

1991-08-01

In order to investigate the feasibility of the domestic passive reactor development, the analysis and evaluation on the development status, technical characteristics, and the safety and economy for the overseas passive reactors were carried out based on the vendor's information. Also the domestic nuclear technology basis was surveyed. The analysis and evaluation of the development status and technical characteristics were performed mainly for the AP-600 developed by Westing house and the SIR of UKAEA. The new design concepts and system characteristics have been evaluated by utilizing EPRI Utility Requirement Documents and Lahmeyer evaluation criteria. Based on this evaluation the recommendable design concepts in each major system were selected. The feasibility for the domestic passive reactor development has focused on the safety, technology and economy aspects, and on the applicability of the existing domestic technology to the design of the passive reactor. And the development plan for the domestic passive reactor was recommended in a step by step way. (Author)

Evaluation of potential blanket concepts for a Demonstration Tokamak Hybrid Reactor

International Nuclear Information System (INIS)

Chapin, D.L.; Chi, J.W.H.; Kelly, J.L.

1978-01-01

An evaluation has been made of several different blanket concepts for use in a near-term Demonstration Tokamak Hybrid Reactor (DTHR), whose main objective would be to produce a significant amount of fissile fuel while demonstrating the feasibility of the tokamak hybrid reactor concept. The desirability of a simple design using proven technology plus a proliferation resistant fuel cycle led to the selection of a low temperature and pressure water-cooled, zircaloy clad ThO 2 blanket concept to breed 233 U. The nuclear performance and thermal-hydraulics characteristics of the blanket were evaluated to arrive at a consistent design. The blanket was found to be feasible for producing a significant amount of fissile fuel even with the limited operating conditions and blanket coverage in the DTHR

Fuel element concept for long life high power nuclear reactors

Science.gov (United States)

Mcdonald, G. E.; Rom, F. E.

1969-01-01

Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

Safety aspects of designs for future light water reactors (evolutionary reactors)

International Nuclear Information System (INIS)

1993-07-01

The main purpose of this document is to describe the major innovations of proposed designs of future light water reactors, to describe specific safety characteristics and safety analysis methodologies, and to give a general overview of the most important safety aspects related to future reactors. The reactors considered in this report are limited to those intended for fixed station electrical power production, excluding most revolutionary concepts. More in depth discussion is devoted to those designs that are in a more advanced state of completion and have been more extensively described and analysed in the open literature. Other designs will be briefly described, as evidence of the large spectrum of new proposals. Some designs are similar; others implement unique features and require specific discussion (not all aspects of designs with unique features are fully discussed in this document). 131 refs, 22 figs

Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1989-01-01

The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-03-01

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

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-01-01

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

Comparison of Design Concepts for SFR under Development

Energy Technology Data Exchange (ETDEWEB)

Suh, Namduk; Choi, Yongwon; Bae, Moohoon; Shin, Andong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-10-15

The goal of ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) with a capacity of 600 MWe is to study the technical demonstration that can be scaled up to commercial reactor. It was expected that the success of ASTRID project could eventually lead to operation of industrial reactor around 2040. On 2012, ASTRID designer has submitted the DOrS (Dossier d’Orientations de Sûreté, Safety Orientation Document) for ASTRID to IRSN and IRSN has issued a report after reviewing the DOrS. The report DOrS itself is not available publicly, intellectual property might be the reason, but the review document of IRSN is open to public, so we can understand the basic concept of ASTRID by IRSN report. The DOrS of ASTRID and the TTR for PGSFR have not the same format and also the same purpose, so it is not easy to compare the two design concepts directly. But, still, we think the concepts could be compared in a very general way. Thus, in this paper we have presented the very short comparison results of the two SFR design. Our opinion after first reviewing the TTR is that the PGSFR needs to be designed in a more systematic way. The requirements are coming basically from the previous document used for SMART licensing and do not show prototype reactor specific characters.

Advanced concept of reduced-moderation water reactor (RMWR) for plutonium multiple recycling

International Nuclear Information System (INIS)

Okubo, T.; Iwamura, T.; Takeda, R.; Yamamoto, K.; Okada, H.

2001-01-01

An advanced water-cooled reactor concept named the Reduced-Moderation Water Reactor (RMWR) has been proposed to attain a high conversion ratio more than 1.0 and to achieve the negative void reactivity coefficient. At present, several types of design concepts satisfying both the design targets have been proposed based on the evaluation for the fuel without fission products and minor actinides. In this paper, the feasibility of the RMWR core is investigated for the plutonium multiple recycling under advanced reprocessing schemes with low decontamination factors as proposed for the FBR fuel cycle. (author)

Conceptual design of the steady state tokamak reactor (SSTR)

International Nuclear Information System (INIS)

Oikawa, A.; Kikuchi, M.; Seki, Y.; Nishio, S.; Ando, T.; Ohara, Y.; Takizuka, Tani, K.; Ozeki, T.; Koizumi, K.; Ikeda, B.; Suzuki, Y.; Ueda, N.; Kageyama, T.; Yamada, M.; Mizoguchi, T.; Iida, F.; Ozawa, Y.; Mori, S.; Yamazaki, S.; Kobayashi, T.; Adachi, H.J.; Shinya, K.; Ozaki, A.; Asahara, M.; Konishi, K.; Yokogawa, N.

1992-01-01

This paper reports that on the basis of a high bootstrap current fraction observation with JT-60, the concept of steady state tokamak reactor , the SSTR, was conceived and was evolved with the design activity of the SSTR at JAERI. Also results of ITER/FER design activities has enhanced the SSTR design. Moreover the remarkable progress of R and D for fusion reactor engineering, especially in the development of superconducting coils and negative ion based NBI at JAERI have promoted the SSTR conceptual design as a realistic power reactor. Although present fusion power reactor designs are currently considered to be too large and costly, results of the SSTR conceptual design suggest that an efficient and promising tokamak reactor will be feasible. The conceptual design of the SSTR provides a realistic reference for a demo tokamak reactor

Design features to achieve defence-in-depth in small and medium sized reactors

International Nuclear Information System (INIS)

Kuznetsov, Vladimir

2009-01-01

Broader incorporation of inherent and passive safety design features has become a 'trademark' of many advanced reactor concepts, including several evolutionary designs and nearly all innovative small and medium sized design concepts. Ensuring adequate defence-in-depth is important for reactors of smaller output because many of them are being designed to allow more proximity to the user, specifically, when non-electrical energy products are targeted. Based on the activities recently performed by the International Atomic Energy Agency, the paper provides a summary description of the design features used to achieve defence in depth in the eleven representative concepts of small and medium sized reactors. (author)

Conceptual designs of tokamak reactor and R D

International Nuclear Information System (INIS)

Fukai, Yuzo; Yamato, Harumi; Sawada, Yoshio

1983-01-01

The conceptual design of both FER (Fusion Experimental Reactor) and R-project is now under way as the new step of JT-60. From the engineering viewpoint, these reactors, requiring D-T operation, have the challenge, such as the handling of tritium and components irradiated by neutron bombardment. Toshiba's design team is participating to these projects in order to realize the reactor and plant concept coping with the above objectives. This paper represents the conceptual design contributions of the FER and R-project as well as R D technology which are now under development, such as tritium handling app aratus, reactor materials, etc. (author)

New reactor concepts; Nieuwe rectorconcepten - nouveaux reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A

1998-11-01

The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost.

Radiotoxicity study of a boiling water reactor core design based on a thorium-uranium fuel concept

International Nuclear Information System (INIS)

Nunez C, A.; Espinosa P, G.

2007-01-01

Full text: The innovative design of a Boiling Water Reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to 233 U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR. A comparison of the toxicity of the spent fuel showed that toxicity is lower in the thorium cycle than other commercial fuels as UO 2 and MOX (uranium and plutonium) in case of the one-through cycle for LWR. (Author)

Evolution of the technical concept of fast reactors. The concept of BREST

International Nuclear Information System (INIS)

Orlov, V.V.

2001-01-01

Having understood that conventional power was limited by available fuel resources, as well as the environmental concern, and willing to use the advantages of defense nuclear power achievements, the development of civil nuclear power was initiated. Scarce supply of uranium has been a matter of concern from the very beginning of nuclear power development, but plutonium produced in the thermal reactors was supposed to be used as fuel for the fast reactors which would not be limited by fuel resources. In order to attain high breeding ratio and high power density, the first generation of fast reactors were designed with sodium coolant, uranium blanket to make up for a decrease in breeding ratio if uranium oxides were used as fuel. Development of nuclear power in the sixties and seventies was followed by stagnation. Lessons learned from a 50-year experience and new conditions set for power industry demand a new concept of fast reactor which would meet a variety of cost-efficiency and safety requirements in their present understanding. Development of fast breeders in Russia began after commissioning of BN-350 and completion of BN-600 design. According to present demands BREST reactors should be designed so as to implement consistently the principles of natural safety without deviation from materials and technology which was proven in defense and civil nuclear power facilities

Development of core design and analyses technology for integral reactor

International Nuclear Information System (INIS)

Zee, Sung Quun; Lee, C. C.; Kim, K. Y.

2002-03-01

In general, small and medium-sized integral reactors adopt new technology such as passive and inherent safety concepts to minimize the necessity of power source and operator actions, and to provide the automatic measures to cope with any accidents. Specifically, such reactors are often designed with a lower core power density and with soluble boron free concept for system simplification. Those reactors require ultra long cycle operation for higher economical efficiency. This cycle length requirement is one of the important factors in the design of burnable absorbers as well as assurance of shutdown margin. Hence, both computer code system and design methodology based on the today's design technology for the current commercial reactor cores require intensive improvement for the small and medium-sized soluble boron free reactors. New database is also required for the development of this type of reactor core. Under these technical requirements, conceptual design of small integral reactor SMART has been performed since July 1997, and recently completed under the long term nuclear R and D program. Thus, the final objectives of this work is design and development of an integral reactor core and development of necessary indigenous design technology. To reach the goal of the 2nd stage R and D program for basic design of SMART, design bases and requirements adequate for ultra long cycle and soluble boron free concept are established. These bases and requirements are satisfied by the core loading pattern. Based on the core loading pattern, nuclear, and thermal and hydraulic characteristics are analyzed. Also included are fuel performance analysis and development of a core protection and monitoring system that is adequate for the soluble boron free core of an integral reactor. Core shielding design analysis is accomplished, too. Moreover, full scope interface data are produced for reactor safety and performance analyses and other design activities. Nuclear, thermal and

Seismic design of reactors in NUCEF

Energy Technology Data Exchange (ETDEWEB)

Kurosaki, Akira [Mitsui Shipbuilding and Engineering Co. Ltd., Tokyo (Japan); Kuchiya, Masao; Yasuda, Naomitsu; Kitanaka, Tsutomu; Ogawa, Kazuhiko; Sakuraba, Koichi; Izawa, Naoki; Takeshita, Isao

1997-03-01

Basic concept and calculation method for the seismic design of the main equipment of the reactors in NUCEF (Nuclear Fuel Cycle Safety Engineering Research Facility) are described with actual calculation examples. The present paper is published to help the seismic design of the equipment and application of the authorization for the design and constructing of facilities. (author)

Reactor physical experimental program EROS in the frame of the molten salt applying reactor concepts development

International Nuclear Information System (INIS)

Hron, Miloslav; Kyncl, Jan; Mikisek, Miroslav

2009-01-01

After the relatively broad program of experimental activities, which have been involved in the complex R and D program for the Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept development in the Czech Republic, there has been a next stage (namely large-scale experimental verification of design inputs by use of MSR-type inserted zones into the existing light water moderated experimental reactor LR-0 called EROS project) started, which will be focused to the experimental verification of the rector physical or neutronic properties of other types of reactor concepts applying molten salts in the role of liquid fuel and/or coolant. This tendency is based on the recently accepted decision of the MSR SSC of GIF to consider for further period of its activity two baseline concepts- fast neutron molten salt reactor non-moderated (FMSR-NM) as a long-term alternative to solid fuelled fast neutron reactors and simultaneously, advanced high temperature reactor (AHTR) with pebble bed type solid fuel cooled by liquid salts. There will be a brief description of the prepared and performed experimental programs in these directions (as well as the preliminary results obtained so far) introduced in the paper. (author)

Conceptual design of the advanced marine reactor MRX

International Nuclear Information System (INIS)

1991-02-01

Design studies on the advanced marine reactors have been done continuously since 1983 at JAERI in order to develop attractive marine reactors for the next generation. At present, two marine reactor concepts are being formulated. One is 100 MWt MRX (Marine Reactor X) for an icebreaker and the other is 300 kWe DRX (Deep-sea Reactor X) for a deep-sea research vessel. They are characterized by an integral type PWR, built-in type control rod drive mechanisms, a water-filled container and a passive decay heat removal system, which realize highly passive safe and compact reactors. This paper is a detailed report including all major results of the MRX design study. (author)

Design characteristics of zero power fast reactor Lasta

International Nuclear Information System (INIS)

Milosevic, M.; Stefanovic, D.; Pesic, M.; Popovic, D.; Nikolic, D.; Antic, D.; Zavaljevski, N.

1987-01-01

The concept, purpose and preliminary design of a zero power fast reactor LASTA are described. The methods of computing the reactor core parameters and reactor kinetics are presented with the basic calculated results and analysis for one selected LASTA configuration. The nominal parameters are determined according to the selected reactor safety criteria and results of calculations. Important aspects related to the overall safety are examined in detail. (author)

Small reactor technical and design characteristics proposed for Indonesia

International Nuclear Information System (INIS)

Nurdin, M.

1992-01-01

A Team for Small Nuclear Electricity Reactor has been formed in Indonesia since June 1990. It is responsible for assessment and design of a small reactor for electricity and/or sea-water desalination. This concept may become a good alternative for power-plants for small islands and for isolated areas in Indonesia, the system should function economically and environmentally sound. In addition to existing concepts, this presentation deals with modifications proposed in improving reliability and safety of reactor operation. For the size of 200 MWth or more (80 MWe or more), the possibility of designing an internal auxiliary heat removal system is discussed, hence there are two separate heat sinks for the core. Future development works for this concept should be directed in expanding their spectrum of utilization and their contribution to the national energy needs. (author). 7 refs., 4 tabs

Plant Control Concept for the Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Kim, Eui Kwang; Kim, S. O.

2010-12-01

A power plant is designed for incorporation into a utility's grid system and follows the load demand through the steam generator, intermediate heat exchanger(IHX), from the nuclear core. During the load-following transients, various plant parameters must be controlled to protect the reactor core and other components in the plant. The purpose of this report is to review design considerations to establish SFR plant control and to design plant control concepts. The governing equations and solution procedure of the computer code to calculate plant temperature conditions during the part-load operation was reviewed and 4 types of plant operation concepts were designed, and the results of the calculations were compared

The dual fluid reactor - a new concept for a highly effective fast reactor

Energy Technology Data Exchange (ETDEWEB)

Huke, A.; Ruprecht, G. [Institut fur Festkorper-Kernphysik gGmbH, Leistikowstr, Berlin (Germany); WeiBbach, D. [Institut fur Festkorper-Kernphysik gGmbH, Leistikowstr, Berlin (Germany); Univ. Szczecin, ul. Wielkopolska, Inst. Fizyki, Wydzial Matematyczno-Fizyczny, Szczecin, (Poland); Gottlieb, S. [Institut fur Festkorper-Kernphysik gGmbH, Leistikowstr, Berlin (Germany); Hussein, A. [Institut fur Festkorper-Kernphysik gGmbH, Leistikowstr, Berlin (Germany); Univ. of Northern British Columbia, Dept. of Physics, Prince George, BC (Canada); Czerski, K. [Institut fur Festkorper-Kernphysik gGmbH, Leistikowstr, Berlin (Germany); Univ. Szczecin, ul. Wielkopolska, Inst. Fizyki, Wydzial Matematyczno-Fizyczny, Szczecin, (Poland)

2014-07-01

The Dual Fluid Reactor, DFR, is a novel concept of a fast heterogeneous nuclear reactor. Its key feature is the employment of two separate liquid cycles, one for fuel and one for the coolant. As opposed to other liquid-fuel concepts like the molten-salt fast reactor (MSFR), in the DFR both cycles can be separately optimized for their respective purpose, leading to advantageous consequences: A very high power density resulting in enormous cost savings, and a highly negative temperature feedback coefficient, enabling a self-regulation without any control rods or mechanical parts in the core. The fuel liquid, an undiluted actinide trichloride (consisting of isotope-purified {sup 37}Cl) in the reference design, circulates at an operating temperature of 1300 K and can be processed on-line in a small internal processing unit utilizing fractionated distillation or electro refining. Medical radioisotopes like Mo-99/Tc-99m are by-products and can be provided right away. In a more advanced design, an actinide metal alloy melt with an appropriately low solidus temperature is well possible which further compactifies the core and allows to further increase the operating temperature due to its high heat conductivity. The best choice for the coolant is pure lead which yields a very hard neutron spectrum. (author)

Packed-fluidized-bed blanket concept for a thorium-fueled commercial tokamak hybrid reactor

International Nuclear Information System (INIS)

Chi, J.W.H.; Miller, J.W.; Karbowski, J.S.; Chapin, D.L.; Kelly, J.L.

1980-09-01

A preliminary design of a thorium blanket was carried out as a part of the Commercial Tokamak Hybrid Reactor (CTHR) study. A fixed fuel blanket concept was developed as the reference CTHR blanket with uranium carbide fuel and helium coolant. A fixed fuel blanket was initially evaluated for the thorium blanket study. Subsequently, a new type of hybrid blanket, a packed-fluidized bed (PFB), was conceived. The PFB blanket concept has a number of unique features that may solve some of the problems encountered in the design of tokamak hybrid reactor blankets. This report documents the thorium blanket study and describes the feasibility assessment of the PFB blanket concept

Evaluation of Metal-Fueled Surface Reactor Concepts

International Nuclear Information System (INIS)

Poston, David I.; Marcille, Thomas F.; Kapernick, Richard J.; Hiatt, Matthew T.; Amiri, Benjamin W.

2007-01-01

Surface fission power systems for use on the Moon and Mars may provide the first use of near-term reactor technology in space. Most near-term surface reactor concepts specify reactor temperatures <1000 K to allow the use of established material and power conversion technology and minimize the impact of the in-situ environment. Metal alloy fuels (e.g. U-10Zr and U-10Mo) have not traditionally been considered for space reactors because of high-temperature requirements, but they might be an attractive option for these lower temperature surface power missions. In addition to temperature limitations, metal fuels are also known to swell significantly at rather low fuel burnups (∼1 a/o), but near-term surface missions can mitigate this concern as well, because power and lifetime requirements generally keep fuel burnups <1 a/o. If temperature and swelling issues are not a concern, then a surface reactor concept may be able to benefit from the high uranium density and relative ease of manufacture of metal fuels. This paper investigates two reactor concepts that utilize metal fuels. It is found that these concepts compare very well to concepts that utilize other fuels (UN, UO2, UZrH) on a mass basis, while also providing the potential to simplify material safeguards issues

Design study of 'HIBLIC-I' reactor cavity

International Nuclear Information System (INIS)

Fujiie, Y.

1984-01-01

A preliminary conceptual design of a reactor cavity for HIBLIC-1, a heavy ion fusion reactor system, was carried out. Design efforts have been concentrated mainly on the feasibility study of the physical scenario adopted and also on the system integration of the structures and components into a compact reactor cavity. The design features of the reactor are a compact reactor cavity, maximum coolant temperature up to 500 deg C, the protection of the sacrificial wall and cavity wall from radiation, the protection of the sacrificial wall from the pressure transient due to rapid heating, the selection of a ferritic steel HT-9 as the structural material and impurity control, and tritium breeding and recovery. The purpose of this paper is to describe the outline of the reactor cavity design of HIBLIC-1. The objectives of the preliminary conceptual design were to propose the idea and concept in order to constitute the physical scenario without contradiction and to find out the critical and fundamental problems to be studied in future. The cavity configuration and dynamics, tritium breeding and radiation damage, the behavior of a structural material in liquid lithium and tritium recovery are reported. (Kako, I.)

Experimental investigation of a directionally enhanced DHX concept for high temperature Direct Reactor Auxiliary Cooling Systems

International Nuclear Information System (INIS)

Hughes, Joel T.; Blandford, Edward D.

2016-01-01

Highlights: • A novel directional heat exchanger design has been developed. • Hydrodynamic tests have been performed on the proposed design. • Heat transfer performance is inferred by hydrodynamic results. • Results are discussed and future work is suggested. - Abstract: The use of Direct Reactor Auxiliary Cooling Systems (DRACSs) as a safety-related decay heat removal system for advanced reactors has developed historically through the Sodium Fast Reactor (SFR) community. Beginning with the EBR-II, DRACSs have been utilized in a large number of past and current SFR designs. More recently, the DRACS has been adopted for Fluoride Salt-Cooled High-Temperature Reactors (FHRs) for similar decay heat removal functions. In this paper we introduce a novel directionally enhanced DRACS Heat Exchanger (DHX) concept. We present design options for optimizing such a heat exchanger so that shell-side heat transfer is enhanced in one primary coolant flow direction and degraded in the opposite coolant flow direction. A reduced-scale experiment investigating the hydrodynamics of a directionally enhanced DHX was built and the data collected is presented. The concept of thermal diodicity is expanded to heat exchanger technologies and used as performance criteria for evaluating design options. A heat exchanger that can perform as such would be advantageous for use in advanced reactor concepts where primary coolant flow reversal is expected during Loss-of-Forced-Circulation (LOFC) accidents where the ability to circulate coolant is compromised. The design could also find potential use in certain advanced Sodium Fast Reactor (SFR) designs utilizing fluidic diode concepts.

Experimental investigation of a directionally enhanced DHX concept for high temperature Direct Reactor Auxiliary Cooling Systems

Energy Technology Data Exchange (ETDEWEB)

Hughes, Joel T.; Blandford, Edward D., E-mail: edb@unm.edu

2016-07-15

Highlights: • A novel directional heat exchanger design has been developed. • Hydrodynamic tests have been performed on the proposed design. • Heat transfer performance is inferred by hydrodynamic results. • Results are discussed and future work is suggested. - Abstract: The use of Direct Reactor Auxiliary Cooling Systems (DRACSs) as a safety-related decay heat removal system for advanced reactors has developed historically through the Sodium Fast Reactor (SFR) community. Beginning with the EBR-II, DRACSs have been utilized in a large number of past and current SFR designs. More recently, the DRACS has been adopted for Fluoride Salt-Cooled High-Temperature Reactors (FHRs) for similar decay heat removal functions. In this paper we introduce a novel directionally enhanced DRACS Heat Exchanger (DHX) concept. We present design options for optimizing such a heat exchanger so that shell-side heat transfer is enhanced in one primary coolant flow direction and degraded in the opposite coolant flow direction. A reduced-scale experiment investigating the hydrodynamics of a directionally enhanced DHX was built and the data collected is presented. The concept of thermal diodicity is expanded to heat exchanger technologies and used as performance criteria for evaluating design options. A heat exchanger that can perform as such would be advantageous for use in advanced reactor concepts where primary coolant flow reversal is expected during Loss-of-Forced-Circulation (LOFC) accidents where the ability to circulate coolant is compromised. The design could also find potential use in certain advanced Sodium Fast Reactor (SFR) designs utilizing fluidic diode concepts.

Very High Efficiency Reactor (VHER) Concepts for Electrical Power Generation and Hydrogen Production

International Nuclear Information System (INIS)

PARMA JR, EDWARD J.; PICKARD, PAUL S.; SUO-ANTTILA, AHTI JORMA

2003-01-01

The goal of the Very High Efficiency Reactor study was to develop and analyze concepts for the next generation of nuclear power reactors. The next generation power reactor should be cost effective compared to current power generation plant, passively safe, and proliferation-resistant. High-temperature reactor systems allow higher electrical generating efficiencies and high-temperature process heat applications, such as thermo-chemical hydrogen production. The study focused on three concepts; one using molten salt coolant with a prismatic fuel-element geometry, the other two using high-pressure helium coolant with a prismatic fuel-element geometry and a fuel-pebble element design. Peak operating temperatures, passive-safety, decay heat removal, criticality, burnup, reactivity coefficients, and material issues were analyzed to determine the technical feasibility of each concept

Summary of trial design of improved marine nuclear reactors

International Nuclear Information System (INIS)

1984-01-01

In order to carry out the research and development of improved marine nuclear reactors, the Japan Nuclear Ship Research and Development Agency decided the project for the purpose in accordance with the procedure of research and development shown by the Nuclear Ship Research and Development Committee of Atomic Energy Commission in December, 1979, and along the basic plan regarding the development of nuclear ships of the Agency decided in February, 1981. As the first step, the Agency has been advancing the research on the design evaluation comprising the trial design and conceptual design to establish the concept of the marine reactor plant with excellent economical efficiency and reliability, which will be developed as the practical plant for future nuclear ships. The trial design started as a three-year project from 1983 is related to a 100 MWt marine reactor, and it is to obtain the concept of improved marine reactors which can be realized after adequate development period based on the pressurized water reactors of separate type, one-body type and semi-one-body type. In this summary, the works carried out in fiscal year 1983 are reported, that is, the design and calculation of the reactor core and the equipment of primary cooling system, and the selection of the required items of research and development. (Kako, I.)

Conceptual design strategy for liquid-metal-wall inertial-fusion reactors

International Nuclear Information System (INIS)

Monsler, M.J.; Meier, W.R.

1981-02-01

The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade provides a basis for formulating a conceptual design strategy for such chambers. Both the attractive and unattractive features of a LMW chamber are enumerated, and a design strategy is formulated which accommodates the engineering constraints while minimizing the liquid-metal flow rate

Conceptual design strategy for liquid-metal-wall inertial-fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Monsler, M.J.; Meier, W.R.

1981-02-01

The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade provides a basis for formulating a conceptual design strategy for such chambers. Both the attractive and unattractive features of a LMW chamber are enumerated, and a design strategy is formulated which accommodates the engineering constraints while minimizing the liquid-metal flow rate.

Space nuclear reactor concepts for avoidance of a single point failure

International Nuclear Information System (INIS)

El-Genk, M. S.

2007-01-01

This paper presents three space nuclear reactor concepts for future exploration missions requiring electrical power of 10's to 100's kW, for 7-10 years. These concepts avoid a single point failure in reactor cooling; and they could be used with a host of energy conversion technologies. The first is lithium or sodium heat pipes cooled reactor. The heat pipes operate at a fraction of their prevailing capillary or sonic limit. Thus, when a number of heat pipes fail, those in the adjacent modules remove their heat load, maintaining reactor core adequately cooled. The second is a reactor with a circulating liquid metal coolant. The reactor core is divided into six identical sectors, each with a separate energy conversion loop. The sectors in the reactor core are neurotically coupled, but hydraulically decoupled. Thus, when a sector experiences a loss of coolant, the fission power generated in it will be removed by the circulating coolant in the adjacent sectors. In this case, however, the reactor fission power would have to decrease to avoid exceeding the design temperature limits in the sector with a failed loop. These two reactor concepts are used with energy conversion technologies, such as advanced Thermoelectric (TE), Free Piston Stirling Engines (FPSE), and Alkali Metal Thermal-to- Electric Conversion (AMTEC). Gas cooled reactors are a better choice to use with Closed Brayton Cycle engines, such as the third reactor concept to be presented in the paper. It has a sectored core that is cooled with a binary mixture of He-Xe (40 gm/mole). Each of the three sectors in the reactor has its own CBC and neutronically, but not hydraulically, coupled to the other sectors

Steady-state operation requirements of tokamak fusion reactor concepts

International Nuclear Information System (INIS)

Knobloch, A.F.

1991-06-01

In the last two decades tokamak conceptual reactor design studies have been deriving benefit from progressing plasma physics experiments, more depth in theory and increasing detail in technology and engineering. Recent full-scale reactor extrapolations such as the US ARIES-I and the EC Reference Reactor study provide information on rather advanced concepts that are called for when economic boundary conditions are imposed. The ITER international reactor design activity concentrated on defining the next step after the JET generation of experiments. For steady-state operation as required for any future commercial tokamak fusion power plants it is essential to have non-inductive current drive. The current drive power and other internal power requirements specific to magnetic confinement fusion have to be kept as low as possible in order to attain a competitive overall power conversion efficiency. A high plasma Q is primarily dependent on a high current drive efficiency. Since such conditions have not yet been attained in practice, the present situation and the degree of further development required are characterized. Such development and an appropriately designed next-step tokamak reactor make the gradual realization of high-Q operation appear feasible. (orig.)

Nuclear design of ISER [intrinsically safe and economical reactor

International Nuclear Information System (INIS)

Yamano, Naoki; Yokoyama, Takashi

1985-01-01

A preliminary core design work on ISER (Intrinsically Safe and Economical Reactor) based on the concept of the PIUS reactor of ASEA-ATOM is performed in order to grasp the characteristics of the reactor core and the fuel management scheme. Certain relations between the fuel specifications and the cycle length are estimated. Items of improvement on the ISER core characteristics and problems to be considered on the nuclear design are presented. Experiments to be considered are also discussed in conjunction with the development of experimental reactor (ISER-E)

Conceptual design study on inertial confinement reactor ''SENRI-II''

International Nuclear Information System (INIS)

Nakamura, N.; Ouura, H.

1983-01-01

Design features of a laser fusion reactor concept SENRI-II are reviewed and discussed. A conceptual design study of the ICF reactor SENRI-II (an advanced design of SENRI-I) has been carried out over 2 years in the Research Committee of ICF Reactors, Institute of Laser Engineering, Osaka University. While the ICF reactor SENRI-I utilized a magnetic field to guide and control an inner liquid lithium flow, SENRI-II is designed to use porous metal as the liquid lithium flow guide. In the design of SENRI-II, a metal porous lithium blanket serves as the protection of a wall against fusion products and as wall per se. Because of the separation of these two functions, a high power density can be attained

Health physics in fusion reactor design

International Nuclear Information System (INIS)

Wong, K.Y.; Dinner, P.J.

1984-06-01

Experience in the control of tritium exposures to workers and the public gained through the design and operation of Ontario Hydro's nuclear stations has been applied to fusion projects and to design studies on emerging fusion reactor concepts. Ontario Hydro performance in occupational tritium exposure control and environmental impact is reviewed. Application of tritium control technologies and dose management methodology during facility design is highlighted

Preliminary design studies on the Broad Application Test Reactor

International Nuclear Information System (INIS)

Terry, W.J.; Terry, W.K.; Ryskamp, J.M.; Jahshan, S.N.; Fletcher, C.D.; Moore, R.L.; Leyse, C.F.; Ottewitte, E.H.; Motloch, C.G.; Lacy, J.M.

1992-08-01

This report describes progress made at the Idaho National Engineering Laboratory during the first three quarters of Fiscal Year (FY) 1992 on the Laboratory-Directed Research and Development (LDRD) project to perform preliminary design studies on the Broad Application Test Reactor (BATR). This work builds on the FY-92 BATR studies, which identified anticipated mission and safety requirements for BATR and assessed a variety of reactor concepts for their potential capability to meet those requirements. The main accomplishment of the FY-92 BATR program is the development of baseline reactor configurations for the two conventional conceptual test reactors recommended in the FY-91 report. Much of the present report consists of descriptions and neutronics and thermohydraulics analyses of these baseline configurations. In addition, we considered reactor safety issues, compared the consequences of steam explosions for alternative conventional fuel types, explored a Molten Chloride Fast Reactor concept as an alternate BATR design, and examined strategies for the reduction of operating costs. Work planned for the last quarter of FY-92 is discussed, and recommendations for future work are also presented

Study on plant concept for gas cooled fast reactor

International Nuclear Information System (INIS)

Moribe, Takeshi; Kubo, Shigenobu; Saigusa, Toshiie; Konomura, Mamoru

2003-05-01

In 'Feasibility Study on Commercialized Fast Reactor Cycle System', technological options including various coolant (sodium, heavy metal, gas, water, etc.), fuel type (MOX, metal, nitride) and output power are considered and classified, and commercialized FBR that have economical cost equal to LWR are pursued. In conceptual study on gas cooled FBR in FY 2002, to identify the prospect of the technical materialization of the helium cooled FBR using coated particle fuel which is an attractive concept extracted in the year of FY2001, the preliminary conceptual design of the core and entire plant was performed. This report summarizes the results of the plant design study in FY2002. The results of study is as follows. 1) For the passive core shutdown equipment, the curie point magnet type self-actuated device was selected and the device concept was set up. 2) For the reactor block, the concept of the core supporting structure, insulators and liners was set up. For the material of the heat resistant structure, SiC was selected as a candidate. 3) For the seismic design of the plant, it was identified that a design concept with three-dimensional base isolation could be feasible taking the severe seismic condition into account. 4) For the core catcher, an estimation of possible event sequences under severe core damage condition was made. A core catcher concept which may suit the estimation was proposed. 5) The construction cost was roughly estimated based on the amount of materials and its dependency on the plant output power was evaluated. The value for a small sized plant exceeds the target construction cost about 20%. (author)

Challenges of designing fusion reactors for remote maintainability

International Nuclear Information System (INIS)

Masson, L.S.

1981-01-01

One of the major problems faced by the fusion community is the development of the high level of reliability required to assure that fusion will be a viable commercial power source. Much of the responsibility for solving this problem falls directly on the designer in developing concepts that have a high level of maintainability for the next generation engineering oriented reactors; and long range, in developing full maintainability for the more complicated commercial concepts with their required high level of on-line time. The near-term challenge will include development of unique design concepts to perform inspection, maintenance, replacement, and testing under the stringent conditions imposed by the next generation engineering oriented machines. The long range challenge will focus on basic design concepts that will enable the full maintainability required by commercial fusion. In addition to the purely technical challenges, the fusion community is also faced with the problem of developing programmatic means to assure that reactor maintenance issues are given proper and timely emphasis as the nuclear phase of fusion is approached

NSSS Component Control System Design of Integral Reactor

International Nuclear Information System (INIS)

Lee, Joon Koo; Kwon, Ho Je; Jeong, Kwong Il; Park, Heui Youn; Koo, In Soo

2005-01-01

MMIS(Man Machine Interface System) of an integral reactor is composed of a Control Room, Plant Protection System, Control System and Monitoring System which are related with the overall plant operation. MMIS is being developed with a new design concept and digital technology to reduce the Human Factor Error and improve the systems' safety, reliability and availability. And CCS(component control system) is also being developed with a new design concept and digital hardware technology A fully digitalized system and design concept are introduced in the NSSS CCS

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Development of mechanical design technology for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Keun Bae; Choi, Suhn; Kim, Kang Soo; Kim, Tae Wan; Jeong, Kyeong Hoon; Lee, Gyu Mahn

1999-03-01

While Korean nuclear reactor strategy seems to remain focused on the large capacity power generation, it is expected that demand of small and medium size reactor will arise for multi-purpose application such as small capacity power generation, co-generation and sea water desalination. With this in mind, an integral reactor SMART is under development. Design concepts, system layout and types of equipment of integral reactor are significantly different from those of loop type reactor. Conceptual design development of mechanical structures of integral reactor SMART is completed through the first stage of the project. Efforts were endeavored for the establishment of design basis and evaluation of applicable codes and standards. Design and functional requirements of major structural components were setup, and three dimensional structural modelling of SMART reactor vessel assembly was prepared. Also, maintenance and repair scheme as well as preliminary fabricability evaluation were carried out. Since small integral reactor technology includes sensitive technologies and know-how's, it is hard to achieve systematic and comprehensive technology transfer from nuclear-advanced countries. Thus, it is necessary to develop the related design technology and to verify the adopted methodologies through test and experiments in order to assure the structural integrity of reactor system. (author)

Development of mechanical design technology for integral reactor

International Nuclear Information System (INIS)

Park, Keun Bae; Choi, Suhn; Kim, Kang Soo; Kim, Tae Wan; Jeong, Kyeong Hoon; Lee, Gyu Mahn

1999-03-01

While Korean nuclear reactor strategy seems to remain focused on the large capacity power generation, it is expected that demand of small and medium size reactor will arise for multi-purpose application such as small capacity power generation, co-generation and sea water desalination. With this in mind, an integral reactor SMART is under development. Design concepts, system layout and types of equipment of integral reactor are significantly different from those of loop type reactor. Conceptual design development of mechanical structures of integral reactor SMART is completed through the first stage of the project. Efforts were endeavored for the establishment of design basis and evaluation of applicable codes and standards. Design and functional requirements of major structural components were set up, and three dimensional structural modelling of SMART reactor vessel assembly was prepared. Also, maintenance and repair scheme as well as preliminary fabricability evaluation were carried out. Since small integral reactor technology includes sensitive technologies and know-how's, it is hard to achieve systematic and comprehensive technology transfer from nuclear-advanced countries. Thus, it is necessary to develop the related design technology and to verify the adopted methodologies through test and experiments in order to assure the structural integrity of reactor system. (author)

Design concept of HYPER (HYbrid Power Extraction Reactor)

International Nuclear Information System (INIS)

Park, Won S.; Song, Tae Y.; Yu, Dong H.; Kim, Chang H.

1999-01-01

Korea Atomic Energy Research Institute (KAERI) has been performing accelerator driven system related research and development called HYPER for the transmutation of nuclear waste and energy production through the transmutation process. Some major design features of HYPER have been developed and employed. On-power fueling concepts are employed to keep system power constant with a minimum variation of accelerator power. A hollow cylinder-type metal fuel is designed for the on-line refueling concept. Pb-Bi is adopted as a coolant and spallation target material. 1 GeV 13 mA proton beam is designed to be provided for HYPER. HYPER is to transmute about 380 kg of TRU a year and produce 1000 MWth power. The support ratio of HYPER for LWR units producing the same power is believed to be 5 to 6. (author)

A new small modular high-temperature gas-cooled reactor plant concept based on proven technology

International Nuclear Information System (INIS)

McDonald, C.F.; Goodjohn, A.J.

1982-01-01

Based on the established and proven high-temperature gas-cooled reactor (HTGR) technologies from the Peach Bottom 1 and Fort St. Vrain utility-operated units, a new small modular HTGR reactor is currently being evaluated. The basic nuclear reactor heat source, with a prismatic core, is being designed so that the decay heat can be removed by passive means (i.e., natural circulation). Although this concept is still in the preconceptual design stage, emphasis is being placed on establishing an inherently safe or benign concept which, when engineered, will have acceptable capital cost and power generation economics. The proposed new HTGR concept has a variety of applications, including electrical power generation, cogeneration, and high-temperature process heat. This paper discusses the simplest application, i.e., a steam Rankine cycle electrical power generating version. The gas-cooled modular reactor concepts presented are based on a graphite moderated prismatic core of low-power density (i.e., 4.1 W/cm 3 ) with a thermal rating of 250 MW(t). With the potential for inherently safe characteristics, a new small reactor could be sited close to industrial and urban areas to provide electrical power and thermal heating needs (i.e., district and space heating). Incorporating a multiplicity of small modular units to provide a larger power output is also discussed. The potential for a small, inherently safe HTGR reactor concept is highlighted

Conceptual design of a commercial accelerator driven thorium reactor

International Nuclear Information System (INIS)

Fuller, C. G.; Ashworth, R. W.

2010-01-01

This paper describes the substantial work done in underpinning and developing the concept design for a commercial 600 MWe, accelerator driven, thorium fuelled, lead cooled, power producing, fast reactor. The Accelerator Driven Thorium Reactor (ADTR TM) has been derived from original work by Carlo Rubbia. Over the period 2007 to 2009 Aker Solutions commissioned this concept design work and, in close collaboration with Rubbia, developed the physics, engineering and business model. Much has been published about the Energy Amplifier concept and accelerator driven systems. This paper concentrates on the unique physics developed during the concept study of the ADTR TM power station and the progress made in engineering and design of the system. Particular attention is paid to where the concept design has moved significantly beyond published material. Description of challenges presented for the engineering and safety of a commercial system and how they will be addressed is included. This covers the defining system parameters, accelerator sizing, core and fuel design issues and, perhaps most importantly, reactivity control. The paper concludes that the work undertaken supports the technical viability of the ADTR TM power station. Several unique features of the reactor mean that it can be deployed in countries with aspirations to gain benefit from nuclear power and, at 600 MWe, it fits a size gap for less mature grid systems. It can provide a useful complement to Generation III, III+ and IV systems through its ability to consume actinides whilst at the same time providing useful power. (authors)

Weapons-grade plutonium dispositioning. Volume 3: A new reactor concept without uranium or thorium for burning weapons-grade plutonium

International Nuclear Information System (INIS)

Ryskamp, J.M.; Schnitzler, B.G.; Fletcher, C.D.

1993-06-01

The National Academy of Sciences (NAS) requested that the Idaho National Engineering Laboratory (INEL) examine concepts that focus only on the destruction of 50,000 kg of weapons-grade plutonium. A concept has been developed by the INEL for a low-temperature, low-pressure, low-power density, low-coolant-flow-rate light water reactor that destroys plutonium quickly without using uranium or thorium. This concept is very safe and could be designed, constructed, and operated in a reasonable time frame. This concept does not produce electricity. Not considering other missions frees the design from the paradigms and constraints used by proponents of other dispositioning concepts. The plutonium destruction design goal is most easily achievable with a large, moderate power reactor that operates at a significantly lower thermal power density than is appropriate for reactors with multiple design goals. This volume presents the assumptions and requirements, a reactor concept overview, and a list of recommendations. The appendices contain detailed discussions on plutonium dispositioning, self-protection, fuel types, neutronics, thermal hydraulics, off-site radiation releases, and economics

Conceptual designs for advanced, high-temperature CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Bushby, S.J. [Atomic Energy of Canada Ltd., Corrosion and Surface Science Branch, Chalk River Laboratories, Chalk River, ON (Canada); Dimmick, G.R. [Atomic Energy of Canada Ltd., Fuel Channel Thermmalhydraulics Branch, Chalk River, ON (Canada); Duffey, R.B. [Atomic Energy of Canada Ltd., Principal Scientist, Chalk River Laboratories, Chalk River, On (Canada); Spinks, N.J. [Atomic Energy of Canada Ltd., Researcher Emeritus, Chalk River Laboratories, Chalk River, ON (Canada); Burrill, K.A. [Atomic Energy of Canada Ltd., Chalk River Laboratories, Chalk River, ON (Canada); Chan, P.S.W. [Atomic Energy of Canada Ltd., Reactor Core Physics Branch, Mississauga, ON (Canada)

2000-07-01

AECL is studying advanced reactor concepts with the aim of significant cost reduction through improved thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, also incorporates enhanced safety features, and flexible, proliferation-resistant fuel cycles, whilst retaining the fundamental design characteristics of CANDU: neutron economy, horizontal fuel channels, and a separate D{sub 2}O moderator that provides a passive heat sink. Where possible, proven, existing components and materials would be adopted, so that 'first-of-a-kind' costs and uncertainties are avoided. Three reactor concepts ranging in output from {approx}375 MW(e) to 1150 MW(e) are described. The modular design of a pressure tube reactor allows the plant size for each concept to be tailored to a given market through the addition or removal of fuel channels. Each concept uses supercritical water as the coolant at a nominal pressure of 25 MPa. Core outlet temperatures range from {approx}400degC to 625degC, resulting in substantial improvements in thermodynamic efficiencies compared to current nuclear stations. The CANDU-X Mark 1 concept is an extension of the present CANDU design. An indirect cycle is employed, but efficiency is increased due to higher coolant temperature, and changes to the secondary side; as well, the size and number of pumps and steam generators are reduced. Safety is enhanced through facilitation of thermo-siphoning of decay heat by increasing the temperature of the moderator. The CANDU-X NC concept is also based on an indirect cycle, but natural convection is used to circulate the primary coolant. This approach enhances cycle efficiency and safety, and is viable for reactors operating near the pseudo-critical temperature of water because of large changes in heat capacity and thermal expansion in that region. In the third concept (CANDUal-X), a dual cycle is employed. Supercritical water exits the core and feeds directly into a very high

Requirements, needs, and concepts for a new broad-application test reactor

International Nuclear Information System (INIS)

Ryskamp, J.M.; Fletcher, C.D.; Denison, A.B.; Liebenthal, J.L.

1992-01-01

For a variety of reasons, including (a) the increasing demands of the 1990s regulatory environment, (b) limited existing test capactiy and capability to satisfy projected future testing missions, and (c) an expected increasing need for nuclear information to support development of advanced reactors, there is a need for requirements and preliminary concepts for a new broad-application test reactor (BATR). These requirements must include consideration not only for a broad range of projected testing missions but also for current and projected regulatory compliance and safety requirements. The requirements will form the basis for development and assessment of preconceptual reactor designs and lead to the identification of key technologies to support the government's long-term strategic and programmatic planning. This paper outlines the need for a new BATR and suggests a few preliminary reactor concepts that can meet that need

Present status of inertial confinement fusion reactor design

International Nuclear Information System (INIS)

Mima, Kunioki; Ido, Shunji; Nakai, Sadao.

1986-01-01

Since inertial nuclear fusion reactors do not require high vacuum and high magnetic field, the structure of the reactor cavity becomes markedly simple as compared with tokamak type fusion reactors. In particular, since high vacuum is not necessary, liquid metals such as lithium and lead can be used for the first wall, and the damage of reactor structures by neutrons can be prevented. As for the core, the energy efficiency of lasers is not very high, accordingly it must be designed so that the pellet gain due to nuclear fusion becomes sufficiently high, and typically, the gain coefficient from 100 to 200 is necessary. In this paper, the perspective of pellet gain, the plan from the present status to the practical reactors, and the conceptual design of the practical reactors are discussed. The plan of fuel ignition, energy break-even and high gain by the implosion mode, of which the uncertain factor due to uneven irradiation and instability was limited to the minimum, was clarified. The scenario of the development of laser nuclear fusion reactors is presented, and the concept of the reactor system is shown. The various types of nuclear fusion-fission hybrid reactors are explained. As for the design of inertial fusion power reactors, the engineering characteristics of the core, the conceptual design, water fall type reactors and DD fuel reactors are discussed. (Kako, I.)

Axial heterogeneous core concept applied for super phoenix reactor

International Nuclear Information System (INIS)

Batista, J.L.; Renke, C.A.C.; Waintraub, M.; Santos Bastos, W. dos; Brito Aghina, L.O. de.

1991-11-01

Always maintaining the current design rules, this paper presents a parametric study on the type of axial heterogeneous core concept (CHA), utilizing a core of fast reactor Super Phenix type, reaching a maximum thermal burnup rate of 150000 M W d/t and being managed in single batch. (author)

Nuclear design of a very-low-activation fusion reactor

International Nuclear Information System (INIS)

Cheng, E.T.; Hopkins, G.R.

1983-06-01

An investigation was conducted to study the nuclear design aspects of using very-low-activation materials, such as SiC, MgO, and aluminum for fusion-reactor first wall, blanket, and shield applications. In addition to the advantage of very-low radioactive inventory, it was found that the very-low-activation fusion reactor can also offer an adequate tritium-breeding ratio and substantial amount of blanket nuclear heating as a conventional-material-structured reactor does. The most-stringent design constraint found in a very-low-activation fusion reactor is the limited space available in the inboard region of a tokamak concept for shielding to protect the superconducting toroidal field coil. A reference design was developed which mitigates the constraint by adopting a removable tungsten shield design that retains the inboard dimensions and gives the same shield performance as the reference STARFIRE tokamak reactor design

Iser: an international inherently safe reactor concept

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki

1988-01-01

Iser is a modular standardised 200-300 MWe power reactor based on the PIUS principle. It differs from PIUS in being simpler, and making full use of existing steel-vessel-based LWR technology. Iser is an inherently safe reactor concept under development in Japan. It is a generic concept, not a patented commodity, and it is expected that an international association to develop the concept will be formed. (U.K.)

Integral Fast Reactor concept

Energy Technology Data Exchange (ETDEWEB)

Till, C.E.; Chang, Y.I.

1986-01-01

The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

Integral Fast Reactor concept

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1986-01-01

The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path

Conceptual design study of small lead-bismuth cooled reactor

International Nuclear Information System (INIS)

Chikazawa, Yoshitaka; Hori, Toru; Kida, Masanori; Konomura, Mamoru

2004-11-01

In phase 2 of the feasibility study of commercialized fast reactor cycle systems of JNC, we make a concept of a small sodium cooled reactor for a power source of a city with various requirements, such as, safety and economical competitiveness. various reactor concepts are surveyed and a tank type reactor whose intermediate heat exchanger and primary main pumps are arranged in series is selected. In this study, a compact long life core and a simple reactor structure designs are pursued. The core type is three regional Zr concentration with one Pu enrichment core, the reactor outlet temperature achieves 550degC and the reactor electric output increases from 150 MWe to 165 MWe. The construction cost is much higher than the economical goal in the case of FOAK. But the construction cost in the case of NOAK is estimated to be 85.6% achieving the economical goal. (author)

New fast reactor installation concept

International Nuclear Information System (INIS)

Anon.

1976-01-01

The large size and complexity of fast reactor installations are emphasised and these difficulties will be increased with the advent of fast reactors of higher power. In this connection a new concept of fast reactor installation is described with a view to reducing the size of the installation and enabling most components, including even the primary vessel, to be constructed within the confines of a workshop. Full constructional details are given. (U.K.)

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

International Nuclear Information System (INIS)

Bjornard, Trond; Hockert, John

2011-01-01

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

Toroidal field coil design concept and structural support system for CTHR

Energy Technology Data Exchange (ETDEWEB)

Chianese, R. B.; Kelly, J. L.; Ruck, G. W.

1980-09-01

The CTHR conceptual design consists of a magnetically confined (tokamak) fusion reactor fitted with a fertile uranium blanket. The fusion driver concept was based on an ignited plasma. All concepts and parameters were selected on the basis that technical feasibility would be achieved by 1995 to assure a viable commercial operation in the early to mid-21st century. The reactor was designed to achieve good fissile fuel production, with electricity production being a second order priority. However, the resulting concepts that evolved were all excellent power producers which significantly improved the economic performance. The subsystems discussed in the following paragraphs provide a background of the application for the TF coil design described in this report.

Toroidal field coil design concept and structural support system for CTHR

International Nuclear Information System (INIS)

Chianese, R.B.; Kelly, J.L.; Ruck, G.W.

1980-09-01

The CTHR conceptual design consists of a magnetically confined (tokamak) fusion reactor fitted with a fertile uranium blanket. The fusion driver concept was based on an ignited plasma. All concepts and parameters were selected on the basis that technical feasibility would be achieved by 1995 to assure a viable commercial operation in the early to mid-21st century. The reactor was designed to achieve good fissile fuel production, with electricity production being a second order priority. However, the resulting concepts that evolved were all excellent power producers which significantly improved the economic performance. The subsystems discussed in the following paragraphs provide a background of the application for the TF coil design described in this report

Development of core design and analyses technology for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

1999-03-01

Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

Development of core design and analyses technology for integral reactor

International Nuclear Information System (INIS)

Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

1999-03-01

Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

Fast reactor core concepts to improve transmutation efficiency

International Nuclear Information System (INIS)

Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

2015-01-01

Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate

Revised design for the Tokamak experimental power reactor

International Nuclear Information System (INIS)

Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

1977-03-01

A new, preliminary design has been identified for the tokamak experimental power reactor (EPR). The revised EPR design is simpler, more compact, less expensive and has somewhat better performance characteristics than the previous design, yet retains many of the previously developed design concepts. This report summarizes the principle features of the new EPR design, including performance and cost

Conceptual core designs for a 1200 MWe sodium cooled fast reactor

International Nuclear Information System (INIS)

Joo, H. K.; Lee, K. B.; Yoo, J. W.; Kim, Y. I.

2008-01-01

The conceptual core design for a 1200 MWe sodium cooled fast reactor is being developed under the framework of the Gen-IV SFR development program. To this end, three core concepts have been tested during the development of a core concept: a core with an enrichment split fuel, a core with a single-enrichment fuel with a region-wise varying clad thickness, and a core with a single-enrichment fuel with non-fuel rods. In order to optimize a conceptual core configuration which satisfies the design targets, a sensitivity study of the core design parameters has been performed. Two core concepts, the core with an enrichment-split fuel and the core with a single-enrichment fuel with a region-wise varying clad thickness, have been proposed as the candidates of the conceptual core for a 1200 MWe sodium cooled fast reactor. The detailed core neutronic, fuel behavior, thermal, and safety analyses will be performed for the proposed candidate core concepts to finalize the core design concept. (authors)

Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

International Nuclear Information System (INIS)

Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

1992-07-01

This report describes the EG ampersand G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options

Design concept of the HPLWR moderator flow path

International Nuclear Information System (INIS)

Koehly, Christina; Schulenberg, Thomas; Starflinger, Joerg

2009-01-01

The latest design concept of the High Performance Light Water Reactor (HPLWR) includes a thermal core in which supercritical water at 25 MPa inlet pressure is heated up from 280degC reactor inlet temperature to 500degC core exit temperature in three steps with intermediate coolant mixing to minimize peak cladding temperatures of the fuel rods. Prior to entering the first fuel assemblies, the coolant is used as moderator in water rods inside assemblies, in the gap volume between assembly boxes, as well as in the surrounding axial or radial reflectors. Even though assembly boxes and moderator rods are designed with a certain thermal insulation, heat is generated in the moderator water or transferred to it from the superheated steam inside assemblies, causing concern of natural convection phenomena with uncontrolled neutronic feedback on the core power distribution. Moreover, bypass flows of the moderator water need to be minimized at any thermal expansion of the reactor internal structures to avoid an unpredictable moderator mass flow. The design concept of the moderator flow path described in this paper is trying to overcome these problems. Downward flow of moderator water is limited to sub-cooled conditions, well below the pseudo-critical point of supercritical water. Dedicated orifices are foreseen to allow later correction of the mass flow split. The sealing concept accounts for larger thermal expansions of reactor components by using C-rings or bellows. A welded construction is preferred wherever possible to minimize leakage. The removable steam plenum is aligned at the extractable steam pipes to minimize thermal displacements at the sealing positions. The paper is showing several design details to illustrate the technical solutions. (author)

Design of the prestressed concrete reactor vessel for gas-cooled heating reactors

International Nuclear Information System (INIS)

Becker, G.; Notheisen, C.; Steffen, G.

1987-01-01

The GHR pebble bed reactor offers a simple, safe and economic possibility of heat generation. An essential component of this concept is the prestressed concrete reactor vessel. A system of cooling pipes welded to the outer surface of the liner is used to transfer the heat from the reactor to the intermediate circuit. The high safety of this vessel concept results from the clear separation of the functions of the individual components and from the design principle of the prestressed conncrete. The prestressed concrete structure is so designed that failure can be reliably ruled out under all operating and accident conditions. Even in the extremely improbable event of failure of all decay heat removal systems when decay heat and accumulated heat are transferred passively by natural convection only, the integrity of the vessel remains intact. For reasons of plant availability the liner and the liner cooling system shall be designed so as to ensure safe elimination of failure over the total operating life. The calculations which were peformed partly on the basis of extremely adverse assumption, also resulted in very low loads. The prestressed concrete vessel is prefabricated to the greatest possible extent. Thus a high quality and optimized fabrication technology can be achieved especially for the liner and the liner cooling system. (orig./HP)

Conceptual design report on advanced marine reactor MRX of Japan

International Nuclear Information System (INIS)

Wang Shengguo

1995-01-01

Design studies on the advanced marine reactors have been done continuously since 1983 at Japan Atomic Energy Institute (JAERI) in order to develop attractive marine reactors for the next generation. At present, two concepts of marine reactor are being formulated. One is 100 MWt MRX (marine Reactor X) for the marine reactor and the other is 150 kWe DRX (Deep Sea-Reactor X) for a deep-sea research vessel. They are characterized by an integral type PWR, built-type control rod drive mechanisms, a water-filled container and a passive decay heat removal system, which realize highly passive safe and compact reactors. The paper is a report about all major results of the MRX design study

Main physics features driving design concept and physics design constraints

International Nuclear Information System (INIS)

Fujisawa, Noboru; Sugihara, Masayoshi; Yamamoto, Shin

1987-07-01

Major physics design philosophies are described, which are essential bases for a plasma design and may have significant impacts on a reactor design concept. Those design philosophies are classified into two groups, physics design drivers and physics design constraints. The design drivers are featured by the fact that a designer is free to choose and the choice may be guided by his opinion, such as ignition, a pulse length, an operation scenario, etc.. The design constraints may follow a physical law, such as plasma confinement, β-limit, density limit, and so on. (author)

An innovative fuel design concept for improved light water reactor performance and safety. Final technical report

International Nuclear Information System (INIS)

Tulenko, J.S.; Connell, R.G.

1995-07-01

Light water reactor (LWR) fuel performance is limited by thermal and mechanical constraints associated with the design, fabrication, and operation of fuel in a nuclear reactor. The purpose of this research was to explore a technique for extending fuel performance by thermally bonding LWR fuel with a non-alkaline liquid metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide (UO 2 ) fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Due to the thermal conductivity of the gas, the gas space thermally insulates the fuel pellets from the reactor coolant outside the fuel rod, elevating the fuel temperatures. Filling the gap between the fuel and clad with a high conductivity liquid metal thermally bonds the fuel to the cladding, and eliminates the large temperature change across the gap, while preserving the expansion and pellet loading capabilities. The resultant lower fuel temperature directly impacts fuel performance limit margins and also core transient performance. The application of liquid bonding techniques to LWR fuel was explored for the purposes of increasing LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) has been developed under the program to analyze the in-reactor performance of the liquid metal bonded fuel. An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of Liquid Metal Bonded LWR fuel

Relevant thermal hydraulic aspects of advanced reactors design: status report

International Nuclear Information System (INIS)

1996-11-01

This status report provides an overview on the relevant thermalhydraulic aspects of advanced reactor designs (e.g. ABWR, AP600, SBWR, EPR, ABB 80+, PIUS, etc.). Since all of the advanced reactor concepts are at the design stage, the information and data available in the open literature are still very limited. Some characteristics of advanced reactor designs are provided together with selected phenomena identification and ranking tables. Specific needs for thermalhydraulic codes together with the list of relevant and important thermalhydraulic phenomena for advanced reactor designs are summarized with the purpose of providing some guidance in development of research plans for considering further code development and assessment needs and for the planning of experimental programs

Review of mirror fusion reactor designs

International Nuclear Information System (INIS)

Bender, D.J.

1977-01-01

Three magnetic confinement concepts, based on the mirror principle, are described. These mirror concepts are summarized as follows: (1) fusion-fission hybrid reactor, (2) tandem mirror reactor, and (3) reversed field mirror reactor

Reactor System Design

International Nuclear Information System (INIS)

Chi, S. K.; Kim, G. K.; Yeo, J. W.

2006-08-01

SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

Analysis of a sustainable gas cooled fast breeder reactor concept

International Nuclear Information System (INIS)

Kumar, Akansha; Chirayath, Sunil S.; Tsvetkov, Pavel V.

2014-01-01

Highlights: • A Thorium-GFBR breeder for actinide recycling ability, and thorium fuel feasibility. • A mixture of 232 Th and 233 U is used as fuel and LWR used fuel is used. • Detailed neutronics, fuel cycle, and thermal-hydraulics analysis has been presented. • Run this TGFBR for 20 years with breeding of 239 Pu and 233 U. • Neutronics analysis using MCNP and Brayton cycle for energy conversion are used. - Abstract: Analysis of a thorium fuelled gas cooled fast breeder reactor (TGFBR) concept has been done to demonstrate the self-sustainability, breeding capability, actinide recycling ability, and thorium fuel feasibility. Simultaneous use of 232 Th and used fuel from light water reactor in the core has been considered. Results obtained confirm the core neutron spectrum dominates in an intermediate energy range (peak at 100 keV) similar to that seen in a fast breeder reactor. The conceptual design achieves a breeding ratio of 1.034 and an average fuel burnup of 74.5 (GWd)/(MTHM) . TGFBR concept is to address the eventual shortage of 235 U and nuclear waste management issues. A mixture of thorium and uranium ( 232 Th + 233 U) is used as fuel and light water reactor used fuel is utilized as blanket, for the breeding of 239 Pu. Initial feed of 233 U has to be obtained from thorium based reactors; even though there are no thorium breeders to breed 233 U a theoretical evaluation has been used to derive the data for the source of 233 U. Reactor calculations have been performed with Monte Carlo radiation transport code, MCNP/MCNPX. It is determined that this reactor has to be fuelled once every 5 years assuming the design thermal power output as 445 MW. Detailed analysis of control rod worth has been performed and different reactivity coefficients have been evaluated as part of the safety analysis. The TGFBR concept demonstrates the sustainability of thorium, viability of 233 U as an alternate to 235 U and an alternate use for light water reactor used fuel as a

On blanket concepts of the Helias reactor

International Nuclear Information System (INIS)

Wobig, H.; Harmeyer, E.; Herrnegger, F.; Kisslinger, J.

1999-07-01

The paper discusses various options for a blanket of the Helias reactor HSR22. The Helias reactor is an upgrade version of the Wendelstein 7-X device. The dimensions of the Helias reactor are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 4.75 T, maximum field 10 T, number of field periods 5, fusion power 3000 MW. The minimum distance between plasma and coils is 1.5 m, leaving sufficient space for a blanket and shield. Three options of a breeding blanket are discussed taking into account the specific properties of the Helias configuration. Due to the large area of the first wall (2600 m 2 ) the average neutron power load on the first wall is below 1 MWm .2 , which has a strong impact on the blanket performance with respect to lifetime and cooling requirements. A comparison with a tokamak reactor shows that the lifetime of first wall components and blanket components in the Helias reactor is expected to be at least two times longer. The blanket concepts being discussed in the following are: the solid breeder concept (HCPB), the dual-coolant Pb-17Li blanket concept and the water-cooled Pb-17Li concept (WCLL). (orig.)

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

Energy Technology Data Exchange (ETDEWEB)

Ingersoll, D.T.

2004-07-29

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

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

International Nuclear Information System (INIS)

Ingersoll, D.T.

2004-01-01

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

The advanced MAPLE reactor concept

International Nuclear Information System (INIS)

Lidstone, R.F.; Lee, A.G.; Gillespie, G.E.; Smith, H.J.

1989-01-01

High-flux neutron sources are continuing to be of interest both in Canada and internationally to support materials testing for advanced power reactors, new developments in extracted-neutron-beam applications, and commercial production of selected radioisotopes. The advanced MAPLE reactor concept has been developed to meet these needs. The advanced MAPLE reactor is a new tank-type D 2 O reactor that uses rodded low-enrichment uranium fuel in a compact annular core to generate peak thermal-neutron fluxes of 1 x 10 19 n·s -1 in a central irradiation rig with a thermal power output of 50 MW. Capital and incremental development costs are minimized by using MAPLE reactor technology to the greatest extent practicable

Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-01

The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

Energy Technology Data Exchange (ETDEWEB)

Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

1998-01-01

To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

Concept and optimization of burning and transmutation reactor in nuclear fuel recycle system

International Nuclear Information System (INIS)

Marsodi; Mulyanto; Kitamoto, Asashi.

1994-01-01

Basic concept of B/T reactor, not only produces thermal energy but also performs burning and/or transmutation of MA and long-lived FPs, was introduced here based on numerical computation model. The advantage of nuclear reaction by thermal or fast neutron was combined conceptually with each other in order to maximize the overall B/T rate obtained by a composite system of fast and thermal reactor. According to the mass balance analysis of B/T reactors with P-T treatment, fast reactor hardened neutron energy may be effective for MA burning. Furthermore, a high flux reactor operated by fast or thermal neutron could be different from a reactor with high B/T rate or high capacity for loading of MA and/or long-lived FPs. The purpose of this study is to make clear the concept and the performance of fast and thermal B/T reactor designed under high neutron utilization for HLW disposal. (author)

SWR 1000: The new boiling water reactor power plant concept

International Nuclear Information System (INIS)

Brettschuh, W.

1999-01-01

Siemens' Power Generation Group (KWU) is currently developing - on behalf of and in close co-operation with the German nuclear utilities and with support from various European partners - the boiling water reactor SWR 1000. This advanced design concept marks a new era in the successful tradition of boiling water reactor technology in Germany and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared to large-capacity nuclear power plants as well as coal-fired stations, while at the same time meeting the highest of safety standards, including control of a core melt accident. This objective is met by replacing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burnup all contribute towards meeting this goal. The design concept fulfils international nuclear regulatory requirements and will reach commercial maturity by the year 2000. (author)

Overall plant concept for a tank-type fast reactor

International Nuclear Information System (INIS)

Yamaki, Hideo; Davies, S.M.; Goodman, L.

1984-01-01

Japanese nuclear industries are expressing interest in the merits of the tank-type FBR as a large plant (demonstration) after JOYO (experimental, in operation) and MONJU (prototype, under construction). In response to this growing interest in a tank-type FBR demonstration plant, Hitachi has initiated a conceptual study of a 1000 MWe tank plant concept in collaboration with GE and Bechtel. Key objectives of this study have been: to select reliable and competitive tank plant concepts, with emphases on a seismic-resistant and compact tank reactor system;to select reliable shutdown heat removal system;and to identify R and D items needed for early 1990s construction. Design goals were defined as follows: capital costs must be less than twice, and as close as practical to 1.5 those of equivalent LWR plants;earthquake resistant structures to meet stringent Japanese seismic conditions must be as simple and reliable as practical;safety must be maintained at LWR-equivalent risks;and R and D needs must be limited to minimum cost for the limited time allowed. This paper summarizes the overall plant concepts with some selected topics, whereas detailed descriptions of the reactor assembly and the layout design are found in separate papers

NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) PROGRAM GRANT NUMBER DE-FG03-00SF22168 TECHNICAL PROGRESS REPORT (Aug 15, 2002 to Nov. 15, 2002) - DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE GENERATION IV REACTOR SYSTEMS

International Nuclear Information System (INIS)

Fred R. Mynatt; Andy Kadak; Marc Berte; Larry Miller; Lawrence Townsend; Martin Williamson; Rupy Sawhney; Jacob Fife

2002-01-01

The objectives of this project are to develop and evaluate nuclear power plant designs and layout concepts to maximize the benefits of compact modular Generation IV reactor concepts including factory fabrication and packaging for optimal transportation and siting. This report covers the ninth quarter of the project. The three reactor concept teams have completed initial plant concept development, evaluation and layout. A significant design effort has proceeded with substantial change and evolution from original ideas. The concepts have been reviewed by the industry participants and improvements have been implemented. The third phase, industrial engineering simulation of reactor fabrication has begun

Kriging-based algorithm for nuclear reactor neutronic design optimization

International Nuclear Information System (INIS)

Kempf, Stephanie; Forget, Benoit; Hu, Lin-Wen

2012-01-01

Highlights: ► A Kriging-based algorithm was selected to guide research reactor optimization. ► We examined impacts of parameter values upon the algorithm. ► The best parameter values were incorporated into a set of best practices. ► Algorithm with best practices used to optimize thermal flux of concept. ► Final design produces thermal flux 30% higher than other 5 MW reactors. - Abstract: Kriging, a geospatial interpolation technique, has been used in the present work to drive a search-and-optimization algorithm which produces the optimum geometric parameters for a 5 MW research reactor design. The technique has been demonstrated to produce an optimal neutronic solution after a relatively small number of core calculations. It has additionally been successful in producing a design which significantly improves thermal neutron fluxes by 30% over existing reactors of the same power rating. Best practices for use of this algorithm in reactor design were identified and indicated the importance of selecting proper correlation functions.

Development of conceptual nuclear design of 10MWt research reactor core

International Nuclear Information System (INIS)

Kim, M. H.; Lim, J. Y.; Win, Naing; Park, J. M.

2008-03-01

KAERI has been devoted to develop export-oriented research reactors for a growing world-wide demand of new research reactor construction. Their ambition is that design of Korean research reactor must be competitive in commercial and technological based on the experience of the HANARO core design concept with thermal power of 30MW. They are developing a new research reactor named Advanced HANARO research Reactor (AHR) with thermal power of 20 MW. KAERI has export records of nuclear technology. In 1954-1967 two series of pool type research reactors based on the Russian design, VVR type and IRT type, have been constructed and commissioned in some countries as well as Russia. Nowadays Russian design is introducing again for export to developing countries such as Union of Myanmar. Therefore the objective of this research is that to build and innovative 10 MW research reactor core design based on the concept of HANARO core design to be competitive with Russian research reactor core design. system tool of HELIOS was used at the first stage in both cases which are research reactor using tubular type fuel assemblies and that reactor using pin type fuel assemblies. The reference core design of first kind of research reactor includes one in-core irradiation site at the core center. The neutron flux evaluations for core as well as reflector region were done through logical consistency of neutron flux distributions for individual assemblies. In order to find the optimum design, the parametric studies were carried out for assembly pitch, active fuel length, number of fuel ring in each assembly and so on. Design result shows the feasibility to have high neutron flux at in-core irradiation site. The second kind of research reactor is used the same kind of assemblies as HANARO and hence there is no optimization about basic design parameters. That core has only difference composition of assemblies and smaller specific power than HANARO. Since it is a reference core at first stage

A novel concept for CRIEC-driven subcritical research reactors

International Nuclear Information System (INIS)

Nieto, M.; Miley, G.H.

2001-01-01

A novel scheme is proposed to drive a low-power subcritical fuel assembly by means of a long Cylindrical Radially-convergent Inertial Electrostatic Confinement (CRIEC) used as a neutron source. The concept is inherently safe in the sense that the fuel assembly remains subcritical at all times. Previous work has been done for the possible implementation of CRIEC as a subcritical assembly driver for power reactors. However, it has been found that the present technology and stage of development of IEC-based neutron sources can not meet the neutron flux requirements to drive a system as big as a power reactor. Nevertheless, smaller systems, such as research and training reactors, could be successfully driven with levels of neutron flux that seem more reasonable to be achieved in the near future by IEC devices. The need for custom-made expensive nuclear fission fuel, as in the case of the TRIGA reactors, is eliminated, and the CRIEC presents substantial advantages with respect to the accelerator-driven subcritical reactors in terms of simplicity and cost. In the present paper, a conceptual design for a research/training CRIEC-driven subcritical assembly is presented, emphasizing the description, principle of operation and performance of the CRIEC neutron source, highlighting its advantages and discussing some key issues that require study for the implementation of this concept. (author)

PRELIMINARY DESIGN OF OSCILLATORY FLOW BIODIESEL REACTOR FOR CONTINUOUS BIODIESEL PRODUCTION FROM JATROPHA TRIGLYCERIDES

Directory of Open Access Journals (Sweden)

AZHARI T. I. MOHD. GHAZI

2008-08-01

Full Text Available The concept of a continuous process in producing biodiesel from jatropha oil by using an Oscillatory Flow Biodiesel Reactor (OFBR is discussed in this paper. It has been recognized that the batch stirred reactor is a primary mode used in the synthesis of biodiesel. However, pulsatile flow has been extensively researcehed and the fundamental principles have been successfully developed upon which its hydrodynamics are based. Oscillatory flow biodiesel reactor offers precise control of mixing by means of the baffle geometry and pulsation which facilitates to continuous operation, giving plug flow residence time distribution with high turbulence and enhanced mass and heat transfer. In conjunction with the concept of reactor design, parameters such as reactor dimensions, the hydrodynamic studies and physical properties of reactants must be considered prior to the design work initiated recently. The OFBR reactor design involves the use of simulation software, ASPEN PLUS and the reactor design fundamentals. Following this, the design parameters shall be applied in fabricating the OFBR for laboratory scale biodiesel production.

Conceptual designs for very high-temperature CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Bushby, S.J.; Dimmick, G.R.; Duffey, R.B. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada)

2000-07-01

Although its environmental benefits are demonstrable, nuclear power must be economically competitive with other energy sources to ensure it retains, or increases, its share of the changing and emerging energy markets of the next decades. In recognition of this, AECL is studying advanced reactor concepts with the goal of significant reductions in capital cost through increased thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, examines concepts for the future, but builds on the success of the current CANDU designs by keeping the same fundamental design characteristics: excellent neutron economy for maximum flexibility in fuel cycle; an efficient heavy-water moderator that provides a passive heat sink under upset conditions; and, horizontal fuel channels that enable on-line refueling for optimum fuel utilization and power profiles. Retaining the same design fundamentals takes maximum advantage of the existing experience base, and allows technological and design improvements developed for CANDU-X to be incorporated into more evolutionary CANDU plants in the short to medium term. Three conceptual designs have been developed that use supercritical water (SCW) as a coolant. The increased coolant temperature results in the thermodynamic efficiency of each CANDU-X concept being significantly higher than conventional nuclear plants. The first concept, CANDU-X Mark 1, is a logical extension of the current CANDU design to higher operating temperatures. To take maximum advantage of the high heat capacity of water at the pseudo-critical temperature, water at nominally 25 MPa enters the core at 310{sup o}C, and exits at {approx}410{sup o}C. The high specific heat also leads to high heat transfer coefficients between the fuel cladding and the coolant. As a result, Zr-alloys can be used as cladding, thereby retaining relatively high neutron economy. The second concept, CANDU-X NC, is aimed at markets that require smaller simpler distributed

Conceptual designs for very high-temperature CANDU reactors

International Nuclear Information System (INIS)

Bushby, S.J.; Dimmick, G.R.; Duffey, R.B.

2000-01-01

Although its environmental benefits are demonstrable, nuclear power must be economically competitive with other energy sources to ensure it retains, or increases, its share of the changing and emerging energy markets of the next decades. In recognition of this, AECL is studying advanced reactor concepts with the goal of significant reductions in capital cost through increased thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, examines concepts for the future, but builds on the success of the current CANDU designs by keeping the same fundamental design characteristics: excellent neutron economy for maximum flexibility in fuel cycle; an efficient heavy-water moderator that provides a passive heat sink under upset conditions; and, horizontal fuel channels that enable on-line refueling for optimum fuel utilization and power profiles. Retaining the same design fundamentals takes maximum advantage of the existing experience base, and allows technological and design improvements developed for CANDU-X to be incorporated into more evolutionary CANDU plants in the short to medium term. Three conceptual designs have been developed that use supercritical water (SCW) as a coolant. The increased coolant temperature results in the thermodynamic efficiency of each CANDU-X concept being significantly higher than conventional nuclear plants. The first concept, CANDU-X Mark 1, is a logical extension of the current CANDU design to higher operating temperatures. To take maximum advantage of the high heat capacity of water at the pseudo-critical temperature, water at nominally 25 MPa enters the core at 310 o C, and exits at ∼410 o C. The high specific heat also leads to high heat transfer coefficients between the fuel cladding and the coolant. As a result, Zr-alloys can be used as cladding, thereby retaining relatively high neutron economy. The second concept, CANDU-X NC, is aimed at markets that require smaller simpler distributed power plants

New concepts for the recovery and isotopic separation of tritium in fusion reactors

International Nuclear Information System (INIS)

Dombra, A.H.; Holtslander, W.J.; Miller, A.I.; Canadian Fusion Fuels Technology Project, Toronto, Ontario)

1986-01-01

New concepts for the recovery of tritium from light water coolant of LiPb blankets, and high-pressure helium coolant of Li-ceramic blankets are introduced. Application of these concepts to fusion reactors is illustrated with conceptual system designs for the anticipated NET blanket requirements. (author)

Conceptual design study of fusion experimental reactor (FY86 FER)

International Nuclear Information System (INIS)

Kobayashi, Takeshi; Yamada, Masao; Mizoguchi, Tadanori

1987-09-01

This report describes the results of the reactor configuration/structure design for the fusion experimental reactor (FER) performed in FY 1986. The design was intended to meet the physical and engineering mission of the next step device which was decided by the subcommittee on the next step device of the nuclear fusion council. The objectives of the design study in FY 1986 are to advance and optimize the design concept of the last year because the recommendation of the subcommittee was basically the same as the design philosophy of the last year. Six candidate reactor configurations which correspond to options C ∼ D presented by the subcommittee were extensively examined. Consequently, ACS reactor (Advanced Option-C with Single Null Divertor) was selected as the reference configuration from viewpoints of technical risks and cost performance. Regarding the reactor structure, the following items were investigated intensively: minimization of reactor size, protection of first wall against plasma disruption, simplification of shield structure, reactor configuration which enables optimum arrangement of poloidal field coils. (author)

Main engineering features driving design concept and engineering design constraints

International Nuclear Information System (INIS)

Saito, Ryusei; Kobayashi, Takeshi; Yamada, Masao

1987-09-01

Major engineering design philosophies are described, which are essential bases for an engineering design and may have significant impacts on a reactor design concept. Those design philosophies are classified into two groups, engineering design drivers and engineering design constraints. The design drivers are featured by the fact that a designer is free to choose and the choice may be guided by his opinion, such as coil system, a mechanical configuration, a tritium breeding scenario, etc.. The design constraints may follow a natural law or engineering limit, such as material strength, coil current density, and so on. (author)

Safety features of the MAPLE-X10 reactor design

International Nuclear Information System (INIS)

Lee, A.G.; Bishop, W.E.; Heeds, W.

1990-09-01

The MAPLE-X10 reactor is a D 2 0-reflected, H 2 0-cooled and -moderated pool-type reactor under construction at the Chalk River Nuclear Laboratories. This 10-MW reactor will produce key medical and industrial radio-isotopes such as 99 Mo, 125 I, and 192 Ir. As the prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor since standards for the licensing of new research reactors have not been developed yet by the licensing authority in Canada

Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Presentations

International Nuclear Information System (INIS)

2013-01-01

The objective of the Technical Meeting is to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials

Investigation on innovative water reactor for flexible fuel cycle (FLWR). (1) Conceptual design

International Nuclear Information System (INIS)

Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiko; Ohnuki, Akira; Iwamura, Takamichi

2005-01-01

A concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI) in order to ensure sustainable energy supply in the future based on the well-experienced Light Water Reactor (LWR). The concept aims at effective and flexible utilization of uranium and plutonium resources through plutonium multiple recycling by two stages. In the first stage, the FLWR core realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming LWR-MOX technologies without significant gaps in technical point of view. The core in the second stage represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the core concepts in both stages utilize the compatible and the same size fuel assemblies, and hence during the reactor operation period, the former concept can proceed to the latter in the same reactor system, corresponding flexibly to the expected change in the future circumstances of natural uranium resource, or establishment of economical reprocessing technology of MOX spent fuel. The FLWR is essentially a BWR-type reactor, and its core design is characterized by use of hexagonal-shaped fuel assemblies with the triangular-lattice fuel rod configuration of highly enriched MOX fuel, control rods with Y-shaped blades, and a short and flat core design. Detailed investigations have been performed on the core design, in conjunction with the other related studies such as on thermal hydraulics in the tight lattice core including experimental activities, and the results obtained so far have shown the proposed concept is feasible and promising. (author)

A compact, inherently safe liquid metal reactor plant concept for terrestrial defense power applications

International Nuclear Information System (INIS)

Magee, P.M.; Dubberley, A.E.; Lutz, D.E.; Palmer, R.S.

1987-01-01

A compact, inherently safe, liquid metal reactor concept based on the GE PRISM innovative LMR design has been developed for terrestrial defense power applications in the 2-50 MWe range. The concept uses a small, sodium-cooled, U-5%Zr metal fueled reactor contained within two redundant steel vessels. The core is designed to operate at a low power density and temperature (925 F) and can operate 30 years without refueling. One two primary coolant loops, depending upon the plant size, transport heat from the core to sodium-to-air, double-wall heat exchangers. Power is produced by a gas turbine operated in a closed ''bottoming'' cycle that employs intercoolers between the compressor stages and a recuperator. Inherent safety is provided by passive means only; operator action is not required to ensure plant safety even for events normally considered Beyond Design Basis Accidents. In addition to normal shutdown heat removal via the sodium-to-air heat exchangers, the design utilizes an inherently passive radiant vessel auxiliary cooling system similar to that designed for PRISM. The use of an air cycle gas turbine eliminates the cost and complexity of the sodium-water reactor pressure relief system required for a steam cycle sodium-cooled reactor

Conceptual design of the Purdue compact torus/passive liner fusion reactor

International Nuclear Information System (INIS)

Terry, W.K.

1981-01-01

This proposal describes a program for the conceptual development of a novel fusion reactor design, the Purdue Compact Torus/Passive Liner Reactor. The key features of the concept are described and a comparison is made with a conventional tokamak

Conceptual design of inherently safe integral reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, J. I.; Chang, M. H.; Lee, D. J. and others

1999-03-01

The design concept of a 300 MWt inherently safe integral reactor(ISIR) for the propulsion of extra large and superhigh speed container ship was developed in this report. The scope and contents of this report are as follows : 1. The state of the art of the technology for ship-mounted reactor 2. Design requirements for ISIR 3. Fuel and core design 4. Conceptual design of fluid system 5. Conceptual design of reactor vessel assembly and primary components 6. Performance analyses and safety analyses. Installation of two ISIRs with total thermal power of 600MWt and efficiency of 21% is capable of generating shaft power of 126,000kW which is sufficient to power a container ship of 8,000TEU with 30knot cruise speed. Larger and speedier ship can be considered by installing 4 ISIRs. Even though the ISIR was developed for ship propulsion, it can be used also for a multi-purpose nuclear power plant for electricity generation, local heating, or seawater desalination by mounting on a movable floating barge. (author)

Nuclear reactor design

CERN Document Server

2014-01-01

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

Divertor remote handling for DEMO: Concept design and preliminary FMECA studies

Energy Technology Data Exchange (ETDEWEB)

Carfora, D., E-mail: dario.carfora@gmail.com [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); ENEA/CREATE/Università degli studi Napoli Federico II, 80125 Napoli (Italy); Di Gironimo, G. [ENEA/CREATE/Università degli studi Napoli Federico II, 80125 Napoli (Italy); Järvenpää, J. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

2015-10-15

Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor mover: hydraulic telescopic boom concept design. • An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • FMECA studies started on the DEMO divertor mover. - Abstract: The paper describes a concept design of a remote handling (RH) system for replacing divertor cassettes and cooling pipes in future DEMO fusion power plant. In DEMO reactor design important considerations are the reactor availability and reliable maintenance operations. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative designs of the end effector to grip and manipulate the divertor cassette are presented in this work. Both concepts are hydraulically actuated, based on ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. Taking advantage of the ITER RH background and experience, the proposed hydraulic RH system is compared with the rack and pinion system currently designed for ITER and is an object of simulations at Divertor Test Platform (DTP2) in VTT's Labs of Tampere, Finland. Pros and cons will be put in evidence.

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)

At-reactor storage concepts criteria for preliminary assessment

International Nuclear Information System (INIS)

Boydston, L.A.

1981-12-01

The licensing, safety, and environmental considerations of four wet and four dry at-reactor storage concepts are presented. Physical criteria for each concept are examined to determine the minimum site and facility requirements which must be met by a utility which desires to expand its at-reactor spent fuel storage capability

Savannah River Site reactor hardware design modification study

International Nuclear Information System (INIS)

Fisher, J.E.

1990-01-01

A study was undertaken to assess the merits of proposed design modifications to the Savannah River Site (SRS) reactors. The evaluation was based on the responses calculated by the RELAP5 systems code to double-ended guillotine break loss-of-coolant-accidents (DEGB LOCAs). The three concepts evaluated were (a) elevated plenum inlet piping with a guard vessel and clamshell enclosures, (b) closure of both rotovalves in the affected loop, and (c) closure of the pump suction valve in the affected loop. Each concept included a fast reactor shutdown (to 65% power in 100 ms) and a 2-s ac pump trip. System recovery potential was evaluated for break locations at the pump suction, the pump discharge, and the plenum inlet. The code version used was RELAP5/MOD2.5 version 3d3, a preliminary version of RELAP5/MOD3. The model was a three-dimensional representation of the K-Reactor water plenum and moderator tank. It included explicit representations of all six loops, which were based on the configuration of L-Reactor. A combination of features is recommended to ensure liquid inventory recovery for all break locations. Valve closure design performance for a break location in the short section of piping between the reactor concrete shield and the pump suction valve would benefit from the clamshell enclosing that section of piping. 7 refs., 10 figs., 2 tabs

Status of conceptual safety design study of Japanese sodium-cooled fast reactor

International Nuclear Information System (INIS)

Kubo, Shigenobu; Kurisaka, Kenichi; Niwa, Hajime; Shimakawa, Yoshio

2005-01-01

In this paper, the current conceptual safety design and related evaluation of Japanese Sodium-cooled Fast Reactor which is studied in the framework of the Feasibility Study (FS) on commercialized Fast Reactor Cycle Systems in Japan are described. The purpose of the safety design is to establish a feasible safety concept of FBR which aims at a sustainable energy source of the next generations. The safety targets and the safety design principle are set aiming at realizing worldwide acceptability of the safety level. The basic safety design concept, which can meet the safety targets, was formulated taking along with the defense-in-depth philosophy as the basic safety design principle. In order to cope with wide range of energy and resource demands, there are some various designs both of oxide and metal fuel for JSFR. Some analytical results of typical design basis events, design extension conditions and core damage frequency estimation show the feasibility of the safety design concept for them. (author)

Safety features of the MAPLE-X10 reactor design

International Nuclear Information System (INIS)

Lee, A.G.; Bishop, W.E.; Heeds, W.

1990-01-01

This paper reports on the MAPLE-X10 reactor D 2 O-reflected, H 2 O-cooled and -moderated pool- type reactor, under construction at the Chalk River Nuclear Laboratories. This 10-MW will produce key medical and industrial radioisotopes such as 99 Mo, 125 I, and 192 Ir. The prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor as standards for the licensing of new research reactors have not been developed by the licensing authority in Canada

Multicriteria selection in concept design of a divertor remote maintenance port in the EU DEMO reactor using an AHP participative approach

Energy Technology Data Exchange (ETDEWEB)

Carfora, D. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Gironimo, G. Di, E-mail: giuseppe.digironimo@unina.it [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Esposito, G. [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Mäkinen, H. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Miccichè, G. [ENEA Brasimone, I:40032 Camugnano (Italy); Mozzillo, R. [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy)

2016-11-15

Highlights: • Concept Studies in Divertor Remote Handling. • Prioritization of concept alternatives. • Comparison and evaluation of product alternatives using AHP. - Abstract: The work behind this paper took place in the Eurofusion remote maintenance system project (WPRM) for the EU Demonstration Fusion Power Reactor (DEMO). Following ITER, the aim of DEMO is to demonstrate the capability of generating several hundreds of MW of net electricity by 2050. The main objective of this paper was the study of the most efficient design of the maintenance port for replacing the divertor cassettes in a Remote Handling (RH) point of view. In DEMO overall design, one important consideration is the availability and short down time operations. The inclination of the divertor port has a very important impact on all the RH tasks such as the design of the divertor mover, the divertor locking systems and the end effectors. The current reference scenario of the EU DEMO foresees a 45° inclined port for the remote maintenance (RM) of the divertor in the lower part of the reactor. Nevertheless, in the optic of the systems engineering (SE) approach, in early concept design phase, all possible configurations shall be taken into account. Even the solutions which seem not feasible at all need to be investigated, because they could lead to new and innovative engineering proposals. The different solutions were compared using an approach based on the Analytic Hierarchy Process (AHP). The technique is a multi-criteria decision making approach in which the factors that are important in making a decision are arranged in a hierarchic structure. The results of these studies show how the application of the AHP improved and focused the selection on the concept which is closer to the requirements arose from technical meetings with the experts of the RH field.

Multicriteria selection in concept design of a divertor remote maintenance port in the EU DEMO reactor using an AHP participative approach

International Nuclear Information System (INIS)

Carfora, D.; Gironimo, G. Di; Esposito, G.; Huhtala, K.; Määttä, T.; Mäkinen, H.; Miccichè, G.; Mozzillo, R.

2016-01-01

Highlights: • Concept Studies in Divertor Remote Handling. • Prioritization of concept alternatives. • Comparison and evaluation of product alternatives using AHP. - Abstract: The work behind this paper took place in the Eurofusion remote maintenance system project (WPRM) for the EU Demonstration Fusion Power Reactor (DEMO). Following ITER, the aim of DEMO is to demonstrate the capability of generating several hundreds of MW of net electricity by 2050. The main objective of this paper was the study of the most efficient design of the maintenance port for replacing the divertor cassettes in a Remote Handling (RH) point of view. In DEMO overall design, one important consideration is the availability and short down time operations. The inclination of the divertor port has a very important impact on all the RH tasks such as the design of the divertor mover, the divertor locking systems and the end effectors. The current reference scenario of the EU DEMO foresees a 45° inclined port for the remote maintenance (RM) of the divertor in the lower part of the reactor. Nevertheless, in the optic of the systems engineering (SE) approach, in early concept design phase, all possible configurations shall be taken into account. Even the solutions which seem not feasible at all need to be investigated, because they could lead to new and innovative engineering proposals. The different solutions were compared using an approach based on the Analytic Hierarchy Process (AHP). The technique is a multi-criteria decision making approach in which the factors that are important in making a decision are arranged in a hierarchic structure. The results of these studies show how the application of the AHP improved and focused the selection on the concept which is closer to the requirements arose from technical meetings with the experts of the RH field.

Core design with respect to the safety concept

International Nuclear Information System (INIS)

Kollmar, W.

1981-01-01

In the present paper the following topics are dealt with: Principles of reactor core design and optimization, fuel management and safety concept for higher cycles and results of risk analyses (e.g. rod ejection, steam line break etc.) (RW)

Preliminary Design of KAIST Micro Modular Reactor with Dry Air Cooling

Energy Technology Data Exchange (ETDEWEB)

Baik, Seung Joon; Bae, Seong Jun; Kim, Seong Gu; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-05-15

KAIST research team recently proposed a Micro Modular Reactor (MMR) concept which integrates power conversion unit (PCU) with the reactor core in a single module. Using supercritical CO{sub 2} as a working fluid of cycle can achieve physically compact size due to small turbomachinery and heat exchangers. The objective of this project is to develop a concept that can operate at isolated area. The design focuses especially on the operation in the inland area where cooling water is insufficient. Thus, in this paper the potential for dry air cooling of the proposed reactor will be examined by sizing the cooling system with preliminary approach. The KAIST MMR is a recently proposed concept of futuristic SMR. The MMR size is being determined to be transportable with land transportation. Special attention is given to the MMR design on the dry cooling, which the cooling system does not depend on water. With appropriately designed air cooling heat exchanger, the MMR can operate autonomously. Two types of air cooling methods are suggested. One is using fan and the other is utilizing cooling tower for the air flow. With fan type air cooling method it consumes about 0.6% of generated electricity from the nuclear reactor. Cooling tower occupies an area of 227 m{sup 2} and 59.6 m in height. This design is just a preliminary estimation of the dry cooling method, and therefore more detailed and optimal design will be followed in the next phase.

Design study of plant system for the fusion experimental reactor (FER)

International Nuclear Information System (INIS)

Iida, Hiromasa; Kuroda, Hideo; Yamada, Masao; Suzuki, Tatsushi; Honda, Tsutomu; Ohmura, Hiroshi; Itoh, Shinichi.

1986-11-01

This report describes design study results of the FER plant system. The purpose of this study is to have an image of the FER plant system as a whole by designing major auxiliary systems, reactor building and maintenance and radwaste desposal systems. The major auxiliary systems include tritium, cooling, evacuation and fueling systems. For these each systems, flowdiagrams are studied and designs of devices and pipings are conducted. In the reactor building design, layout of the above auxiliary systems in the building is studied with careful zoning concept by the radiation level. Structural integrity of the reactor building is also studied including seismic analysis. In the design of the maintenance and radwaste system flowdiagram of failed reactor components is developed and transfer vehicles and buildings are designed. Finally assuming JAERI Naka site as the reactor site layout of the whole FER plant system is developed. (author)

The Integral Fast Reactor concept

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1986-01-01

The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes the key features and potential advantages of the IFR concept, its technology development status, fuel cycle economics potential, and its future development path

Once-through cycle, supercritical-pressure light water cooled reactor concept

Energy Technology Data Exchange (ETDEWEB)

Oka, Y.; Koshizuka, S. [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab

2001-07-01

Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

Once-through cycle, supercritical-pressure light water cooled reactor concept

International Nuclear Information System (INIS)

Oka, Y.; Koshizuka, S.

2001-01-01

Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

Knowledge gaps in economic analyses of advanced reactor concepts

International Nuclear Information System (INIS)

Moore, M.; Pencer, J.; Leung, L.K.H.; Sadhankar, R.

2014-01-01

The development of next generation nuclear systems is predicated on improvement in sustainability, safety, proliferation resistance and economics. The economic assessment of the reactor concept is required as early as in the concept development stage. The Generation IV International Forum (GIF) has developed a methodology for economic assessment of the Generation IV (GEN-IV) nuclear energy systems. The GIF economics methodology was used for the assessment of one of the reactor concepts for the Super-Critical Water-cooled Reactors (SCWR), namely the European pressure-vessel type concept referred to as the High Performance Light Water Reactor (HPLWR). The economic analysis involved studying the sensitivity of two main economic indicators, namely, the Levelized Unit Electricity Cost (LUEC) and the Total Capital Investment Cost (TCIC). The knowledge gaps in estimating the capital costs and fuel costs, as well as the uncertainties in other cost parameters affecting the economic assessment of the nuclear energy system in the concept development stage are presented. (author)

Concept of an accelerator-driven subcritical research reactor within the TESLA accelerator installation

International Nuclear Information System (INIS)

Pesic, Milan; Neskovic, Nebojsa

2006-01-01

Study of a small accelerator-driven subcritical research reactor in the Vinca Institute of Nuclear Sciences was initiated in 1999. The idea was to extract a beam of medium-energy protons or deuterons from the TESLA accelerator installation, and to transport and inject it into the reactor. The reactor core was to be composed of the highly enriched uranium fuel elements. The reactor was designated as ADSRR-H. Since the use of this type of fuel elements was not recommended any more, the study of a small accelerator-driven subcritical research reactor employing the low-enriched uranium fuel elements began in 2004. The reactor was designated as ADSRR-L. We compare here the results of the initial computer simulations of ADSRR-H and ADSRR-L. The results have confirmed that our concept could be the basis for designing and construction of a low neutron flux model of the proposed accelerator-driven subcritical power reactor to be moderated and cooled by lead. Our objective is to study the physics and technologies necessary to design and construct ADSRR-L. The reactor would be used for development of nuclear techniques and technologies, and for basic and applied research in neutron physics, metrology, radiation protection and radiobiology

Conceptual design study of Fusion Experimental Reactor (FY87FER)

International Nuclear Information System (INIS)

Miki, Nobuharu; Iida, Fumio; Wachi, Yoshihiro; Toyoda, Katsuyoshi; Hashizume, Takashi; Konno, Masayuki.

1988-06-01

This report describes the FER magnet design which was conducted last year (1987). Based on a large uncertainty of the physics assumption, two sets of FER concepts have been developed. One is based on the best existing physics data bases and another is based on rather conservative physics bases. In the magnet design, the improvements of superconducting magnet design were investigated to reduce the reactor size and to realize higher reactor-core performance. In addition, we studied several critical technical issues that affect the magnet design specification. (author)

Status of advanced technology and design for water cooled reactors: Light water reactors

International Nuclear Information System (INIS)

1988-10-01

Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

Blanket concepts for the ARIES commercial tokamak reactor study

International Nuclear Information System (INIS)

Grotz, S.P.; Ghoniem, N.M.; Hasan, M.Z.; Martin, R.C.; Najmabadi, F.; Sharafat, S.; Hua, T.; Sze, D.K.; Cheng, E.T.; Creedon, R.L.; Wong, C.P.C.; Herring, J.S.; Klein, A.; Snead, L.; Steiner, D.

1989-01-01

The ARIES study is a 3-year effort, started in 1988, exploring the potential of the tokamak to be an attractive and competitive commercial power reactor. Several different versions of the tokamak are being considered, combining different levels of extrapolations in physics and engineering databases. The first version studied in detail, ARIES-I, combines present-day physics (with minimal extrapolation) with aggressive engineering technology such as very high-field, superconducting magnets and low-activation silicon carbide composite materials. The ARIES-I version is designed to meet acceptable safety and environmental criteria. In particular, achieving a passively safe concept that meets Class-C waste disposal is one of the high leverage items in the design. This paper summarizes the scoping analysis and engineering design of the ARIES-I fusion-power-core subsystems. The ARIES-I design is a 1000 MW e power reactor, operating at steady state in the 1 st stability regime and uses a high magnetic field. Typical operating parameters of the ARIES-I strawman design are listed

Fusion reactor design and technology 1986. V. 1

International Nuclear Information System (INIS)

1987-01-01

The first volume of the Proceedings of the Fourth Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology organized by the IAEA (Yalta, 26 May - 6 June 1986) includes 36 papers devoted to the following topics: fusion programmes (3 papers), tokamaks (15 papers), non-tokamak reactors and open systems (9 papers), inertial confinement concepts (5 papers), fission-fusion hybrids (4 papers). Each of these papers has a separate abstract. Refs, figs and tabs

Comparison of three ICF reactor designs

International Nuclear Information System (INIS)

Hogan, W.J.

1984-01-01

Three concepts for inertial confinement fusion (ICF) reactors are described and compared with each other, and with magnetic fusion and fission reactors on the basis of environmental impact, safety and efficiency. The critical technical developments of each concept are described. The three concepts represent alternative development paths for inertial fusion

SCW Pressure-Channel Nuclear Reactor Some Design Features

Science.gov (United States)

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

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

Universally applicable design concept of stably controlling an HTGR-hydrogen production system

International Nuclear Information System (INIS)

Hada, Kazuhiko; Shibata, Taiju; Nishihara, Tetsuo; Shiozawa, Shusaku

1996-01-01

An HTGR-hydrogen production system should be designed to have stable controllability because of a large difference in thermal dynamics between reactor and hydrogen production system and such a control design concept should be universally applicable to a variety of hydrogen production processes by the use of nuclear heat from HTGR. A transient response analysis of an HTGR-steam reforming hydrogen production system showed that a steam generator installed in a helium circuit for cooling the nuclear reactor provides stable controllability of the total system, resulting in avoiding a reactor scram. A survey of control design-related characteristics among several hydrogen production processes revealed the similarity of endothermic chemical reactions by the use of high temperature heat and that steam is required as a reactant of the endothermic reaction or for preheating a reactant. Based on these findings, a system design concept with stable controllability and universal applicability was proposed to install a steam generator as a downstream cooler of an endothermic reactor in the helium circuit of an HTGR-hydrogen production system. (author)

Design and development of small and medium integral reactor core

International Nuclear Information System (INIS)

Zee, Sung Quun; Chang, M. H.; Lee, C. C.; Song, J. S.; Cho, B. O.; Kim, K. Y.; Kim, S. J.; Park, S. Y.; Lee, K. B.; Lee, C. H.; Chun, T. H.; Oh, D. S.; In, W. K.; Kim, H. K.; Lee, C. B.; Kang, H. S.; Song, K. N.

1997-07-01

Recently, the role of small and medium size integral reactors is remarkable in the heat applications rather than the electrical generations. Such a range of possible applications requires extensive used of inherent safety features and passive safety systems. It also requires ultra-longer cycle operations for better plant economy. Innovative and evolutionary designs such as boron-free operations and related reactor control methods that are necessary for simple reactor system design are demanded for the small and medium reactor (SMR) design, which are harder for engineers to implement in the current large size nuclear power plants. The goals of this study are to establish preliminary design criteria, to perform the preliminary conceptual design and to develop core specific technology for the core design and analysis for System-integrated Modular Advanced ReacTor (SMART) of 330 MWt power. Based on the design criteria of the commercial PWR's, preliminary design criteria will be set up. Preliminary core design concept is going to be developed for the ultra-longer cycle and boron-free operation and core analysis code system is constructed for SMART. (author). 100 refs., 40 tabs., 92 figs

Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

International Nuclear Information System (INIS)

1986-10-01

This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc

Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

Energy Technology Data Exchange (ETDEWEB)

1986-10-01

This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc. (C-E).

Challenges of designing fusion reactors for remote maintainability

International Nuclear Information System (INIS)

Mason, L.S.

1981-01-01

One of the major problems faced by the fusion community is the development of the high level of reliability required to assure that fusion will be a viable commercial power source. Much of the responsibility for solving this problem falls directly on the designer in developing concepts that have a high level of maintainability. The problems are both near-term, in developing maintainability for next generation engineering oriented reactors; and long range, in developing full maintainability for the more commercial concepts with their required high level of on-line time. The near-time challenge will include development of unqiue design concepts to perform inspection, maintenance, replacement, and testing under the stringent conditions imposed by the next generation engineering oriented machines. The long range challenge will focus on basic design concepts that will enable the full mainatability required by commerical fusion

Design of a boiling water reactor core based on an integrated blanket-seed thorium-uranium concept

International Nuclear Information System (INIS)

Nunez-Carrera, Alejandro; Francois, Juan Luis; Martin-del-Campo, Cecilia; Espinosa-Paredes, Gilberto

2005-01-01

This paper is concerned with the design of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material in an integrated blanket-seed (BS) assembly. The integrated BS concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned out in a once-through cycle. The idea behind the lattice design is to use the thorium conversion capability in a BWR spectrum, taking advantage of the 233 U build-up. A core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235 U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the fuel assembly

A conceptual design strategy for liquid-metal-wall inertial fusion reactors

International Nuclear Information System (INIS)

Monsler, M.J.; Meier, W.R.

1981-01-01

The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade are reviewed from the perspective of formulating a conceptual design strategy for such chambers. The basis for the design strategy is set by enumerating both the attractive and unattractive features of a LMW chamber. Past concepts are then reviewed to identify conceptual design approaches and physical configurations that enhance the positive aspects and minimize the negative aspects. A detailed description of the engineering considerations is given, including such topics as the selection of a liquid metal, control of radiation damage, selection of structural material, control of tritium breeding and extraction, control of wall stress, and designing for a given rep-rate. Finally, a design strategy is formulated which accomodates the engineering constraints while minimizing the liquid-metal flow rate. (orig.)

Experimental investigation of the MSFR molten salt reactor concept

Energy Technology Data Exchange (ETDEWEB)

Yamaji, Bogdan; Aszodi, Attila [Budapest University of Technology and Economics (Hungary). Inst. of Nuclear Techniques

2014-11-15

In the paper experimental modelling and investigation of the MSFR concept will be presented. MSFR is a homogeneous, single region liquid fuelled fast reactor concept. In case of molten salt reactors the core neutron flux and fission distribution is determined by the flow field through distribution and transport of fissile material and delayed neutron precursors. Since the MSFR core is a single region homogeneous volume without internal structures, it is a difficult task to ensure stable flow field, which is strongly coupled to the volumetric heat generation. These considerations suggest that experimental modelling would greatly help to understand the flow phenomena in such geometry. A scaled and segmented experimental mock-up of MSFR was designed and built in order to carry out particle image velocimetry measurements. Basic flow behaviour inside the core region can be investigated and the measurement data can also provide resource for the validation of computational fluid dynamics models. Measurement results of steady state conditions will be presented and discussed.

Fast mixed spectrum reactor concept

International Nuclear Information System (INIS)

Kouts, H.J.C.; Fischer, G.J.; Cerbone, R.J.

1979-04-01

The Fast Mixed Spectrum Reactor is a highly promising concept for a fast reactor with improved features of proliferation resistance, and excellent utilization of uranium resources. In technology, it can be considered to be a branch of fast breeder development, though its operation and implications are different from those of FBR'S in important respects. Successful development programs are required in several areas to bring FMSR to reality, but the payoff from a successful program can be high

Application of probabilistic risk assessment to advanced liquid metal reactor designs

International Nuclear Information System (INIS)

Carroll, W.P.; Temme, M.I.

1987-01-01

The United States Department of Energy (US DOE) has been active in the development and application of probabilistic risk assessment methods within its liquid metal breeder reactor development program for the past eleven years. These methods have been applied to comparative risk evaluations, the selection of design features for reactor concepts, the selection and emphasis of research and development programs, and regulatory discussions. The application of probabilistic methods to reactors which are in the conceptual design stage presents unique data base, modeling, and timing challenges, and excellent opportunities to improve the final design. We provide here the background and insights on the experience which the US DOE liquid metal breeder reactor program has had in its application of probabilistic methods to the Clinch River Breeder Reactor Plant project, the Conceptual Design State of the Large Development Plant, and updates on this design. Plans for future applications of probabilistic risk assessment methods are also discussed. The US DOE is embarking on an innovative design program for liquid metal reactors. (author)

Remote maintenance design for Fusion Experimental Reactor (FER)

International Nuclear Information System (INIS)

Tachikawa, K.; Iida, H.; Nishio, S.; Tone, T.; Aota, T.; Iwamoto, T.; Niikura, S.; Nishizawa, H.

1984-01-01

Design of Fusion Experimental Reactor, FER, has been conducted by Japan Atomic Energy Research Institute (JAERI) since 1981. Two typical reactors can be classified in general from the viewpoints of remote maintenance among four design concepts of FER. In the case of the type 1 FER, the torus module consists of shield structure and blanket, and the connective joints between toruses provided at the outer region of the reactor. As for the type 2 FER, the shield structure is joined with the vacuum cryostat, and only the blanket module is allowed to move, but connection between toruses are located in the inner region of the reactor. Comparing type 1 with type 2 FER, this paper describes on the remote maintenance of FER including reactor configurations, work procedures, remote systems/equipments, repairing facility and future R and D problems. Reviewing design studies and investigation for the existing robotics technologies, R and D for FER remote maintenance technology should be performed under the reasonable long-term program. The main items of remote technology required to start urgently are multi-purpose manipulator system with performance of dextrousity, tele-viewing system which reduces operator fatigue and remote tests for commercially available components

Design and analysis on super-critical water cooled power reactors

International Nuclear Information System (INIS)

Ishiwatari, Yuki

2005-01-01

The Super-Critical Water Cooled Power Reactors (SCPR) is cooled by 25 MPa supercritical water of 280degC at reactor inlet and greater than 500degC at reactor outlet and directly connected with turbine/generators with high energy conversion efficiency. This corresponds to the deletion of recirculation system and steam-water separation system of BWR type reactors or of pressurizer and steam generator of PWR type reactors. In addition to the design study of the university of Tokyo, technology development of the SCPR for practical use has started under the collaboration of industry and academia since 2000. Mockup single tube and bundle tests for heat transfer/fluid flow characteristics of the design have been conducted with 3D heat transfer analysis. Materials compatible with coolant conditions for fuel cans and reactor internals are also assessed. Overall evaluation of the reactor concept is under way. (T. Tanaka)

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

A conceptual design of LIB fusion reactor: UTLIF(2)

International Nuclear Information System (INIS)

Madarame, Haruki; Kondo, Shunsuke; Iwata, Shuichi; Oka, Yoshiaki; Miya, Kenzo.

1984-01-01

UTLIF(2) is a conceptual design study on a light ion beam driven fusion reactor based on a concept of rod-bundle blanket. Survivability and maintainability of the first wall and the blanket are regarded as of major importance in the design. The blanket rod is composed of a thick tube which has enough stiffness, a thin wrapping wall which receives high heat flux, and liquid lithium which breeds tritium and removes generated heat. The rod can be pulled out from the outside of the reactor vessel, hence the replacement is very easy. Nuclear and thermal analysis have been made and the performance of the reactor has been shown to be satisfactory. (author)

Conceptual design of Fusion Experimental Reactor (FER)

International Nuclear Information System (INIS)

Tone, T.; Fujisawa, N.

1983-01-01

Conceptual design studies of the Fusion Experimental Reactor (FER) have been performed. The FER has an objective of achieving selfignition and demonstrating engineering feasibility as a next generation tokamak to JT-60. Various concepts of the FER have been considered. The reference design is based on a double-null divertor. Optional design studies with some attractive features based on advanced concepts such as pumped limiter and RF current drive have been carried out. Key design parameters are; fusion power of 440 MW, average neutron wall loading of 1MW/m 2 , major radius of 5.5m, plasma minor radius of 1.1m, plasma elongation of 1.5, plasma current of 5.3MA, toroidal beta of 4%, toroidal field on plasma axis of 5.7T and tritium breeding ratio of above unity

Savannah River Site reactor hardware design modification study

International Nuclear Information System (INIS)

Fisher, J.E.

1990-03-01

A study was undertaken to assess the merits of proposed design modifications to the SRS reactors. The evaluation was based on the responses calculated by the RELAP5 systems code to double-ended guillotine break loss-of-coolant-accidents (DEGB LOCAs). The three concepts evaluated were (a) elevated plenum inlet piping with a guard vessel and clamshell enclosures, (b) closure of both rotovalves in the affected loop, and (c) closure of the pump suction valve in the affected loop. Each concept included a fast reactor shutdown (to 65% power in 100 ms) and a 2-s ac pump trip. For the elevated piping design, system recovery was predicted for breaks in the plenum inlet or pump suction piping; response to the pump discharge break location did not show improvement compared to the present system configuration. The rotovalve closure design improved system response to plenum inlet or pump discharge breaks; recovery was not predicted for pump suction breaks. The pump suction valve closure design demonstrated system recovery for all break locations downstream of the valve. A combination of features is recommended to ensure liquid inventory recovery for all break locations. The elevated piping design performance during pump discharge breaks would be improved with addition of a dc pump trip in the affected loop. Valve closure design performance for a break location in the short section of piping between the reactor concrete shield and the pump suction valve would benefit from the clamshell enclosing that section of piping. 12 refs., 10 figs., 2 tabs

The concept of the innovative power reactor

Directory of Open Access Journals (Sweden)

Sang Won Lee

2017-10-01

Full Text Available The Fukushima accident reveals the vulnerability of existing active nuclear power plant (NPP design against prolonged loss of external electricity events. The passive safety system is considered an attractive alternative to cope with this kind of disaster. Also, the passive safety system enhances both the safety and the economics of NPPs. The adoption of a passive safety system reduces the number of active components and can minimize the construction cost of NPPs. In this paper, reflecting on the experience during the development of the APR+ design in Korea, we propose the concept of an innovative Power Reactor (iPower, which is a kind of passive NPP, to enhance safety in a revolutionary manner. The ultimate goal of iPower is to confirm the feasibility of practically eliminating radioactive material release to the environment in all accident conditions. The representative safety grade passive system includes a passive emergency core cooling system, a passive containment cooling system, and a passive auxiliary feedwater system. Preliminary analysis results show that these concepts are feasible with respect to preventing and/or mitigating the consequences of design base accidents and severe accidents.

Safety concerns and suggested design approaches to the HTGR Reformer process concept

Energy Technology Data Exchange (ETDEWEB)

Green, R.C.

1981-09-01

This report is a safety review of the High Temperature Gas-Cooled Reactor Reformer Application Study prepared by Gas-Cooled Reactor Associates (GCRA) of La Jolla, California. The objective of this review was to identify safety concerns and suggests design approaches to minimize risk in the High Temperature Gas-Cooled Reactor Reformer (HTGR-R) process concept.

Safety concerns and suggested design approaches to the HTGR Reformer process concept

International Nuclear Information System (INIS)

Green, R.C.

1981-09-01

This report is a safety review of the High Temperature Gas-Cooled Reactor Reformer Application Study prepared by Gas-Cooled Reactor Associates (GCRA) of La Jolla, California. The objective of this review was to identify safety concerns and suggests design approaches to minimize risk in the High Temperature Gas-Cooled Reactor Reformer (HTGR-R) process concept

Evaluation of the trial design studies for an advanced marine reactor, (3)

International Nuclear Information System (INIS)

Ambo, Noriaki; Yokomura, Takeyoshi.

1988-03-01

JAERI have carried out four core designs for three different type reactors (Semi-Integrated, Integrated and Integrated (self-pressured) type reactors), as the trial designs of an Advanced Marine Reactor for three years (1983 ∼ 1985). This report describes the result of comparison and studies of the core specific characteristics of these four cores, which include core concept, specifications, core life, specific power density, burn-up, reactivity control and etc. In conclusion, it was found that the Integrated type reactor core and the Semi-Integrated type reactor core designs satisfy the conditions of long core life (four years), high specific power density (50 ∼ 61 kw/l) and high burn-up (30,000 ∼ 32,000 MWD/t), so these two cores will be optimum designs based on the present technologies. (author)

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)

Design of controller for control rod of research reactors

International Nuclear Information System (INIS)

Abou-Zaid, R.M.F.M

2008-01-01

Designing and testing digital control system for any nuclear research reactor can be costly and time consuming. In this thesis, a rapid, low-cost proto typing and testing procedure for digital controller design is proposed using the concept of Hardware-In-The-Loop (HIL). Some of the control loop components are real hardware components and the others are simulated. First, the whole system is modeled and tested by Real-Time Simulation (RTS) using conventional simulation techniques such as MATLAB / SIMULINK. Second the Hardware-in-the-loop simulation is tested using Real-Time Windows Target in MATLAB and Visual C ++ . The control parts are included as hardware components which are the reactor control rod and its drivers. Three kinds of controllers are studied, Proportional-Derivative (PD), Proportional-Integral-Derivative (PID) and Fuzzy controller. An experimental setup for the hardware used in HIL concept for the control of the nuclear research reactor has been realized. Experimental results are obtained and compared with the simulation results. The experimental results indicate the validation of HIL method in this domain.

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)

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1985-01-01

The Fusion Experimental Reactor (FER) being developed at JAERI as a next generation tokamak to JT-60 has a major mission of realizing a self-ignited long-burning DT plasma and demonstrating engineering feasibility. During FY82 and FY83 a comprehensive and intensive conceptual design study has been conducted for a pulsed operation FER as a reference option which employs a conventional inductive current drive and a double-null divertor. In parallel with the reference design, studies have been carried out to evaluate advanced reactor concepts such as quasi-steady state operation and steady state operation based on RF current drive and pumped limiter, and comparative studies for single-null divertor/pumped limiter. This report presents major results obtained primarily from FY83 design studies, while the results of FY82 design studies are described in previous references (JAERI-M 83-213--216). (author)

The near boiling reactor: design of a small nuclear reactor for extending the operational envelope of the Victoria Class Submarine

International Nuclear Information System (INIS)

Cole, C.; Bonin, H.

2005-01-01

A small, inherently safe nuclear reactor that will provide enough power to maintain the hotel load of the Victoria Class Submarine and extend her operational envelope, has been conceptually designed. The final reactor concept, named the Near Boiling (NB) Reactor, employs TRISO fuel particles in Zirconium cladded fuel rods. The reactor is light water moderated and cooled. The core life is specifically designed to coincide with the refit cycle of the Victoria Class Submarine. The reactor employs a simple and reliable control and shut down system that requires little intervention on the part of the submarine's crew. Also, a kinetic model is developed that demonstrates the inherent safety features of the reactor during several accident scenarios. (author)

The near boiling reactor: design of a small nuclear reactor for extending the operational envelope of the Victoria Class Submarine

Energy Technology Data Exchange (ETDEWEB)

Cole, C.; Bonin, H. [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)]. E-mail: chris.cole@rmc.ca; bonin-h@rmc.ca

2005-07-01

A small, inherently safe nuclear reactor that will provide enough power to maintain the hotel load of the Victoria Class Submarine and extend her operational envelope, has been conceptually designed. The final reactor concept, named the Near Boiling (NB) Reactor, employs TRISO fuel particles in Zirconium cladded fuel rods. The reactor is light water moderated and cooled. The core life is specifically designed to coincide with the refit cycle of the Victoria Class Submarine. The reactor employs a simple and reliable control and shut down system that requires little intervention on the part of the submarine's crew. Also, a kinetic model is developed that demonstrates the inherent safety features of the reactor during several accident scenarios. (author)

Design Studies for a Multiple Application Thermal Reactor for Irradiation Experiments (MATRIX)

Energy Technology Data Exchange (ETDEWEB)

Pope, Michael A.; Gougar, Hans D.; Ryskamp, J. M.

2015-03-01

The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Should unforeseen circumstances lead to the decommissioning of ATR, the U.S. Government would be left without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. A survey was conducted in order to catalogue the anticipated needs of potential customers. Then, concepts were evaluated to fill the role for this reactor, dubbed the Multi-Application Thermal Reactor Irradiation eXperiments (MATRIX). The baseline MATRIX design is expected to be capable of longer cycle lengths than ATR given a particular batch scheme. The volume of test space in In-Pile-Tubes (IPTs) is larger in MATRIX than in ATR with comparable magnitude of neutron flux. Furthermore, MATRIX has more locations of greater volume having high fast neutron flux than ATR. From the analyses performed in this work, it appears that the lead MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design is developed further.

Safety aspects of the US advanced LMR [liquid metal reactor] design

International Nuclear Information System (INIS)

Pedersen, D.R.; Gyorey, G.L.; Marchaterre, J.F.; Rosen, S.

1989-01-01

The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. This paper discusses the US regulatory framework for design of an ALMR, safety aspects of the IFR program at ANL, the IFR fuel cycle and actinide recycle, and the ALMR plant design program at GE. 6 refs., 5 figs

Status of small reactor designs without on-site refuelling

International Nuclear Information System (INIS)

2007-01-01

There is an ongoing interest in member states in the development and application of small and medium sized reactors (SMRs). In the near term, most new NPPs are likely to be evolutionary designs building on proven systems while incorporating technological advances and often the economics of scale, resulting from the reactor outputs of up to 1600 MW(e). For the longer term, the focus is on innovative designs aiming to provide increased benefits in the areas of safety and security, non-proliferation, waste management, resource utilization and economy, as well as to offer a variety of energy products and flexibility in design, siting and fuel cycle options. Many innovative designs are reactors within the small-to-medium size range, having an equivalent electric power less than 700 MW(e) or even less than 300 MW(e). A distinct trend in design and technology development, accounting for about half of the SMR concepts developed worldwide, is represented by small reactors without on-site refuelling. Such reactors, also known as battery-type reactors, could operate without reloading and shuffling of fuel in the core over long periods, from 5 to 25 years and beyond. Upon the advice and with the support of IAEA member states, within its Programme 1 'Nuclear Power, Fuel Cycle, and Nuclear Science', the IAEA provides a forum for the exchange of information by experts and policy makers from industrialized and developing countries on the technical, economic, environmental, and social aspects of SMRs development and implementation in the 21st century, and makes this information available to all interested Member States by producing status reports and other publications dedicated to advances in SMR technology. The objective of this report is to provide Member States, including those just considering the initiation of nuclear power programmes and those already having practical experience in nuclear power, with a balanced and objective information on important development trends and

Substantiation of physical concepts of fast reactors in Russia: experience and prospects

Energy Technology Data Exchange (ETDEWEB)

Alekseev, P.N. [Russian Research Center ' Kurchatov Institute' (RRC KI), 1, Kurchatov Sq., Moscow, 123182 (Russian Federation); Vasiliev, B.A. [Experimental Design Bureau of Machine Building (OKBM) 15, Burnakovskiy Pr., N. Novgorod, 603074 (Russian Federation); Kormilitsyn, M.V. [State Scientific Center of Russian Federation - Research Institute of Atomic Reactors (NIIAR) Dimitrovgrad-10, Ulianovsk Reg., 433510 (Russian Federation); Lopatkin, A.V. [N.A. Dollezhal Research and Development Institute of Power Engineering (NIKIET) 2/8, M. Krasnoselskaya Str., Moscow, 107140 (Russian Federation); Seleznev, E.F. [All-Russian Research Institute for Nuclear Power Plant Operation (VNIIAES) 25, Ferganskaya, Moscow, 109507 (Russian Federation); Khomyakov, Yu.S.; Tsybulia, A.M. [State Scientific Center of the Russian Federation - A. I. Leypunsky Institute for Physics and Power Engineering (SSC RF- IPPE) 1, Bondarenko Sq., Obninsk, Kaluga Reg., 249033 (Russian Federation); Tocheny, L.V. [International Science and Technology Center (ISTC) 32-34 Krasnoproletarskaya Ulitsa, Moscow, 127473 (Russian Federation)

2008-07-01

The fast reactor concept in Russia has accumulated unique experience, since its advent in the 1950's and up to the present, from the creation of the first experimental installation BR-1, experimental reactors BR-5 and BOR-60, the pilot industrial reactors BN-350 in Kazakhstan and up to the BN-600 at Beloyarsk Atomic Power Station. Investigations on the first experimental installations BR-1 and BR-5/-10 proved the propriety of the idea that it is possible to create nuclear reactors that can produce more nuclear fuel than they consume, i.e. the idea of breeding. The architecture of such reactors was also designed, producing a current leader among fast reactors with sodium coolant and oxide uranium-plutonium fuel. Operational experience of BOR-60, BN-350 and, particularly, BN-600 confirmed the engineering and technical feasibility of the concept of fast reactors, the possibility for its realization both for power production and for certain other purposes as well, such as desalinisation of sea water (BN-350) and for radionuclide production (BN-350, BN-600), and it enabled the development and verification of different models, computer methods and codes. The paper presents a review of experience in the creation of plants with fast reactors, scientific research on these installations, principal results, the current status of experimental data analysis, and prospective directions in the development of fast reactors and the corresponding experimental basis in Russia. (authors)

STARFIRE remote maintenance and reactor facility concept

International Nuclear Information System (INIS)

Graumann, D.W.; Field, R.E.; Lutz, G.R.; Trachsel, C.A.

1981-01-01

A total remote maintenance facility has been designed for all equipment located within the reactor building and hot cell, although operational flexibility has been provided by design of the reactor shielding such that personnel access into the reactor building within 24 hours after reactor shutdown is possible. The reactor design permits removal and replacement of all components if necessary, however, the vacuum pumps, isolation valves and blanket require scheduled, routine maintenance. Reactor scheduled maintenance does not dominate annual plant downtime, therefore, several scheduled operations can be added without affecting reactor availability. The maintenance facilities consist of the reactor building, the hot cell, the reactor service area and the remote maintenance control room. The reactor building contains the reactor, selected support system modules, and required maintenance equipment. The reactor and the support systems are maintained with (1) equipment that is mounted on a monorail system; (2) overhead cranes; and (3) bridge-mounted electromechanical manipulators. The hot cell is located outside of the reactor building to localize contamination products and permit independent operation. An equipment air lock connects the reactor building to the hot cell

Concept of passive safe small reactor for distributed energy supply system

International Nuclear Information System (INIS)

Ishida, Toshihisa; Nakajima, Nobuya; Sawada, Ken-ichi; Yoritsune, Tsutomu; Shimada, Shoichiro; Nakano, Yoshihiro; Yonomoto, Taisuke; Takahashi, Hiroki

2003-01-01

This paper presents a concept of a Passive Safe Small Reactor for Distributed energy supply system (PSRD). The PSRD is an integrated-type PWR with reactor thermal power of 100 to 300 MW aimed at supplying electricity, district heating, etc. In design of the PSRD, high priority is laid on enhancement of safety as well as improvement of economy. Safety is enhanced by the following means: i) Extreme reduction of pipes penetrating the reactor vessel, by limiting to only those of the steam, the feed water and the safety valves, ii) Adoption of the water filled containment and the passive safety systems with fluid driven by natural circulation force, and iii) Adoption of the in-vessel type control rod drive mechanism, accompanying a passive reactor shut-down. To comply with a severe operation condition of PSRD, material of the ball bearing with graphite retainer has been selected by test. For improvement of economy, simplification of the reactor system and long operation of the core are achieved. Optimization of core design concerning the burnable poison ensures the burn-up of 28 GWd/t for low enriched UO 2 fuel rods. (author)

Conceptual design of D-3He FRC reactor 'ARTEMIS'

International Nuclear Information System (INIS)

Momota, H.; Ishida, A.; Kohzaki, Y.

1991-07-01

A comprehensive design study of the D- 3 He fueled field-reversed configuration (FRC) reactor 'ARTEMIS' is carried out for the purpose of proving its attractive characteristics and clarifying the critical issues for a commercial fusion reactor. The FRC burning plasma is stabilized and sustained in a steady equilibrium by means of a preferential trapping of D- 3 He fusion-produced energetic protons. A novel direct energy converter for 15MeV protons is also presented. On the bases of a consistent scenario of the fusion plasma production and simple engineering, a compact and simple reactor concept is presented. The design of the D- 3 He FRC power plant definitely offers the most attractive prospect for energy development. It is environmentally acceptable in view of radio-activity and fuel resources; and the estimated cost of electricity is low compared to a light water reactor. Critical issues concerning physics or engineering for the development of the D- 3 He FRC reactor are clarified. (author)

Development of fluid system design technology for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, D. J.; Chang, M. H.; Kang, D. J. and others

1999-03-01

This study presents the technology development of the system design concepts of SMART, a multi-purposed integral reactor with enhanced safety and operability, for use in diverse usages and applications of the nuclear energy. This report contains the following; - Design characteristics - Performance and safety related design criteria - System description: Primary system, Secondary system, Residual heat removal system, Make-up system, Component cooling system, Safety system - Development of design computer code: Steam generator performance(ONCESG), Pressurizer performance(COLDPZR), Steam generator flow instability(SGINS) - Development of component module and modeling using MMS computer code - Design calculation: Steam generator thermal sizing, Analysis of feed-water temperature increase at a low flow rate, Evaluation of thermal efficiency in the secondary system, Inlet orifice throttling coefficient for the prevention of steam generator flow instability, Analysis of Nitrogen gas temperature in the pressurizer during heat-up process, evaluation of water chemistry and erosion etc. The results of this study can be utilized not only for the foundation technology of the next phase basic system design of the SMART but also for the basic model in optimizing the system concepts for future advanced reactors. (author)

Target/blanket conceptual design for the Los Alamos ATW concept

International Nuclear Information System (INIS)

Ames, K.; Cappiello, M.; Ireland, J.; Sapir, J.; Farnum, G.

1992-01-01

The Los Alamos Accelerator Transmutation of Waste (ATW) concept has many potential applications that include defense waste transmutation, defense material production (i.e., tritium and 238 Pu), and the transmutation of hazardous nuclear wastes from commercial nuclear reactors (fission products and actinides). A more advanced long-term Los Alamos effort is investigating the potential of an accelerator- driven system to produce fission energy with a minimal nuclear waste stream. All applications employ a high-energy (800- to 1600-MeV), high-current (25--250 mA) proton linear accelerator as the driver. In this report, we discuss only the target/blanket conceptual design for the commercial nuclear waste application. A conceptual design for the target/blanket of the Los Alamos ATW concept has been presented. The neutronics, mechanical design, and heat transfer have been investigated in some detail for the base-case design. Much more work needs to be done, but at this point it appears that the design is feasible and will approach the design goal of supporting two commercial power reactors with each target/blanket module

A preliminary conceptual design study for Korean fusion DEMO reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Keeman, E-mail: kkeeman@nfri.re.kr [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Kim, Hyoung Chan; Oh, Sangjun; Lee, Young Seok; Yeom, Jun Ho; Im, Kihak; Lee, Gyung-Su [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Neilson, George; Kessel, Charles; Brown, Thomas; Titus, Peter [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States)

2013-10-15

Highlights: ► Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. ► Present design guidelines and requirements of Korean DEMO reactor. ► Present a preliminary design of TF (toroidal field) and CS (central solenoid) magnet. ► Present a preliminary result of the radial build scheme of Korean DEMO reactor. -- Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb{sub 3}Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters.

A HUMAN AUTOMATION INTERACTION CONCEPT FOR A SMALL MODULAR REACTOR CONTROL ROOM

Energy Technology Data Exchange (ETDEWEB)

Le Blanc, Katya; Spielman, Zach; Hill, Rachael

2017-06-01

Many advanced nuclear power plant (NPP) designs incorporate higher degrees of automation than the existing fleet of NPPs. Automation is being introduced or proposed in NPPs through a wide variety of systems and technologies, such as advanced displays, computer-based procedures, advanced alarm systems, and computerized operator support systems. Additionally, many new reactor concepts, both full scale and small modular reactors, are proposing increased automation and reduced staffing as part of their concept of operations. However, research consistently finds that there is a fundamental tradeoff between system performance with increased automation and reduced human performance. There is a need to address the question of how to achieve high performance and efficiency of high levels of automation without degrading human performance. One example of a new NPP concept that will utilize greater degrees of automation is the SMR concept from NuScale Power. The NuScale Power design requires 12 modular units to be operated in one single control room, which leads to a need for higher degrees of automation in the control room. Idaho National Laboratory (INL) researchers and NuScale Power human factors and operations staff are working on a collaborative project to address the human performance challenges of increased automation and to determine the principles that lead to optimal performance in highly automated systems. This paper will describe this concept in detail and will describe an experimental test of the concept. The benefits and challenges of the approach will be discussed.

An energy amplifier fluidized bed nuclear reactor concept

International Nuclear Information System (INIS)

Sefidvash, F.; Seifritz, W.

2001-01-01

The concept of a fluidized bed nuclear reactor driven by an energy amplifier system is described. The reactor has promising characteristics of inherent safety and passive cooling. The reactor can easily operate with any desired spectrum in order to be a plutonium burner or have it operate with thorium fuel cycle. (orig.) [de

The integral fast reactor (IFR) concept: Physics of operation and safety

International Nuclear Information System (INIS)

Wade, D.C.; Chang, Y.I.

1987-01-01

The IFR concept employs a pool layout, a U/Pu/Zr metal alloy fuel and a closed fuel cycle based on pyrometallurgical reprocessing and injection casting refabrication. The reactor physics issues of designing for inherent safety and for a closed fissile self-sufficient integral fuel cycle with uranium startup and potential actinide transmutation are discussed

The integral fast reactor (IFR) concept: physics of operation and safety

International Nuclear Information System (INIS)

Wade, D.C.; Chang, Y.I.

1987-01-01

The IFR concept employs a pool layout, a U/Pu/Zr metal alloy fuel and a closed fuel cycle based on pyrometallurgical reprocessing and injection casting refabrication. The reactor physics issues of designing for inherent safety and for a closed fissile self-sufficient integral fuel cycle with uranium startup and potential actinide transmutation are discussed

Conceptual design of blanket structures for fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-03-01

Conceptual design study for in-vessel components including tritium breeding blanket of FER has been carried out. The objective of this study is to obtain the engineering and technological data for selecting the reactor concept and for its construction by investigating fully and broadly. The design work covers in-vessel components (such as tritium breeding blanket, first wall, shield, divertor and blanket test module), remote handling system and tritium system. The designs of those components and systems are accomplished in consideration of their accomodation to whole reactor system and problems for furthur study are clarified. (author)

JSFR design progress related to development of safety design criteria for Generation IV sodium-cooled fast reactors. (1) Overview

International Nuclear Information System (INIS)

Kamide, Hideki; Ando, Masato; Ito, Takaya

2015-01-01

JAEA, JAPC and MFBR have been conducting design study for the Japan Sodium-cooled Fast Reactor (JSFR), which is a design concept aiming at future commercial use as sustainable electric power source. As the result of the design study and R and D activity related the innovative technologies incorporated in the design in the Fast Reactor Cycle Technology Development (FaCT) project up to 2010, basic design concept of JSFR was established and its development process to the commercialization including construction and operation of a demonstration version of JSFR was outlined. JSFR is a looptype next generation sodium-cooled fast reactor (SFR), which is aiming at achieving development targets of Generation IV reactors concerning sustainability, safety and reliability, economics and proliferation resistance and physical protection by introducing the innovative technologies such as shortened high-chromium steel piping. The output power is assumed for the design study as 1,500 MWe for the commercial version and 750 MWe for the demonstration version. In FaCT phase I up to 2010, in order to evaluate feasibility to achieve the development targets, the design study has been conducted on the main components and systems. Since 2011, in order to contribute to the development of safety design criteria (SDC) and safety design guideline (SDG), which include the lessons learned from the TEPCO's Fukushima Dai-ichi nuclear power plants accident, in the frame work of Generation IV International Forum (GIF), the design study is focusing on the design measures against severe external events such as earthquake and tsunami. At the same time, the design study is going into detail and paying much attention to the maintenance and repair to make surer its feasibility. This paper summarizes the design concept of the demonstration version of JSFR in which progress of design work was incorporated for the safety issues on SDC and SDG of a SFR. (author)

First wall and blanket design for the STARFIRE commercial tokamak power reactor

International Nuclear Information System (INIS)

Morgan, G.D.; Trachsel, C.A.; Cramer, B.A.; Bowers, D.A.; Smith, D.L.

1979-01-01

The first wall and blanket design concepts being evaluated for the STARFIRE commercial tokamak reactor study are presented. The two concepts represent different approaches to the mechanical design of a tritium breeding blanket using the reference materials options. Each concept has a separate ferritic steel first wall cooled by heavy water (D 2 O), and a ferritic steel blanket with solid lithium oxide breeder cooled by helium. A separate helium purge system is used in both concepts to extract tritium. The two concepts are compared and relative advantages and disadvantages for each are discussed

Advanced reactor concepts and safety

International Nuclear Information System (INIS)

Lipsett, J.J.

1988-06-01

The need for some consistency in the terms used to describe the evolution of methods for ensuring the safety of nuclear reactors has been identified by the IAEA. This is timely since there appears to be a danger that the precision of many valuable words is being diluted and that a new jargon may appear that will confuse rather than aid the communication of important but possibly diverse philosophies and concepts. Among the difficulties faced by the nuclear industry is promoting and gaining a widespread understanding of the risks actually posed by nuclear reactors. In view of the importance of communication to both the public and to the technical community generally, the starting point for the definition of terms must be with dictionary meanings and common technical usage. The nuclear engineering community should use such words in conformance with the whole technical world. This paper addresses many of the issues suggested in the invitation to meet and also poses some additional issues for consideration. Some examples are the role of the operator in either enhancing or degrading safety and how the meaning or interpretation of the word 'safety' can be expected to change during the next few decades. It is advantageous to use criteria against which technologies and ongoing operating performance can be judged provided that the criteria are generic and not specific to particular reactor concepts. Some thoughts are offered on the need to frame the criteria carefully so that innovative solutions and concepts are fostered, not stifled

Containment design, performance criteria and research needs for advanced reactor designs

International Nuclear Information System (INIS)

Bagdi, G.; Ali, S.; Costello, J

2004-01-01

This paper points out some important shifts in the basic expectations in the performance requirements for containment structures and discusses the areas where the containment structure design requirements and acceptance criteria can be integrated with ultimate test based insights. Although there has not been any new reactor construction in the United States for over thirty years, several designs of evolutionary and advanced reactors have already been certified. Performance requirements for containment structures under design basis and severe accident conditions and explicit consideration of seismic margins have been used in the design certification process. In the United States, the containment structure design code is the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NE-Class MC for the steel containment and Section III, Division 2 for reinforced and prestressed concrete reactor vessels and containments. This containment design code was based on the early concept of applying design basis internal pressure and associated load combinations that included the operating basis and safe shutdown earthquake ground motion. These early design criteria served the nuclear industry and the regulatory authorities in maintaining public health and safety. However, these early design criteria do not incorporate the performance criteria related to containment function in an integrated fashion. Research in large scale model testing of containment structures to failure from over pressurization and shake table testing using simulated ground motion, have produced insights related to failure modes and material behavior at failure. The results of this research provide the opportunity to integrate these observations into design and acceptance criteria. This integration process would identify 'gaps' in the present knowledge and future research needs. This knowledge base is important for gleaning risk-informed insights into

Stationary liquid fuel fast reactor SLFFR – Part I: Core design

Energy Technology Data Exchange (ETDEWEB)

Jing, T.; Yang, G.; Jung, Y.S.; Yang, W.S., E-mail: yang494@purdue.edu

2016-12-15

Highlights: • An innovative fast reactor concept SLFFR based on liquid metal fuel is proposed for TRU burning. • A compact core design of 1000 MWt SLFFR is developed to achieve a zero conversion ratio and passive safety. • The core size and the control requirement are significantly reduced compared to the conventional solid fuel reactor with same conversion ratio. - Abstract: For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named the stationary liquid fuel fast reactor (SLFFR) has been proposed based on a stationary molten metallic fuel. A compact core design of a 1000 MWt SLFFR has been developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches have been adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses have been performed to evaluate the steady-state performance characteristics. The analysis results indicate that the SLFFR of a zero TRU conversion ratio is feasible while satisfying the conservatively imposed thermal design constraints. A theoretical maximum TRU consumption rate of 1.01 kg/day is achieved with uranium-free fuel. Compared to the solid fuel reactors with the same TRU conversion ratio, the core size and the reactivity control requirement are reduced significantly. The primary and secondary control systems provide sufficient shutdown margins, and the calculated reactivity feedback coefficients show that the prompt fuel expansion coefficient is sufficiently negative.

Preliminary structural evaluations of the STAR-LM reactor vessel and the support design

International Nuclear Information System (INIS)

Koo, Gyeong-Hoi; Sienicki, James J.; Moisseytsev, Anton

2007-01-01

In this paper, preliminary structural evaluations of the reactor vessel and support design of the STAR-LM (The Secure, Transportable, Autonomous Reactor - Liquid Metal variant), which is a lead-cooled reactor, are carried out with respect to an elevated temperature design and seismic design. For an elevated temperature design, the structural integrity of a direct coolant contact to the reactor vessel is investigated by using a detail structural analysis and the ASME-NH code rules. From the results of the structural analyses and the integrity evaluations, it was found that the design concept of a direct coolant contact to the reactor vessel cannot satisfy the ASME-NH rules for a given design condition. Therefore, a design modification with regards to the thermal barrier is introduced in the STAR-LM design. For a seismic design, detailed seismic time history response analyses for a reactor vessel with a consideration of a fluid-structure interaction are carried out for both a top support type and a bottom support type. And from the results of the hydrodynamic pressure responses, an investigation of the minimum thickness design of the reactor vessel is tentatively carried out by using the ASME design rules

Design criteria for the electrical system in advanced passive reactors. Special features of the AP-600 Reactor

International Nuclear Information System (INIS)

Moraleda Lopez, A.

1997-01-01

The design of the electrical system of an Passive Advanced Reactor is determined by the concept of passive actuation of safety systems, simplification of process systems and optimisation of equipment performance. The system that results from these criteria is very different to those designed for present plants. The main differences are: No class 1E alternating current systems No emergency diesel generators Fewer safety and non-safety class electricity consumers System for continuous monitoring of battery status Use of electronic speed regulators for reactor feedwater pump motors Outsite battery backup safety power supply Motor-operated valves are the only safety electrical actuators Portable power supply for post 72 hour equipment This paper develops these concepts and applies them to the AP-600 project and describes the electrical system of this type of plant. (Author)

Reactor physics challenges in GEN-IV reactor design

Energy Technology Data Exchange (ETDEWEB)

Driscoll, Michael K.; Hejzlar, Pavel [Massachusetts Institute of Technology, MA (United States)

2005-02-15

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

Reactor physics challenges in GEN-IV reactor design

International Nuclear Information System (INIS)

Driscoll, Michael K.; Hejzlar, Pavel

2005-01-01

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources

Safety design concept and analysis for the upgrading JRR-3

International Nuclear Information System (INIS)

Onishi, N.; Isshiki, M.; Takahashi, H.; Takayanagi, M.

1990-01-01

The Research Reactor No.3 (JRR-3) is under reconstruction for upgrading. This paper describes the safety design concepts of the architectural and engineering design, anticipated operational transients and accident conditions which are the postulated initiating events for the safety evaluation, and the safety criteria of the upgraded JRR-3. The safety criteria are defined taking into account those of Light Water Reactors and the characteristics of the research reactor. Using the example of the safety analysis, this paper describes analytical results of a reactivity insertion by removal of in-core irradiation samples, a pipeline break at the primary coolant loop and flow blockage to a coolant channel, which are the severest postulated initiating events of the JRR-3

The design rationale of the Integral Fast Reactor (IFR)

International Nuclear Information System (INIS)

Wade, D.C.; Hill, R.N.

1997-01-01

The Integral Fast Reactor (IFR) concept has been developed over the last ten years to provide technical solutions to perceptual concerns associated with nuclear power. Beyond the traditional advanced reactor objectives of increased safety, improved economy and more efficient fuel utilization, the IFR is designed to simplify waste disposal and increase resistance to proliferation. Only a fast reactor with an efficient recycle technology can provide for total consumption of actinides. The basic physics governing reactor design dictates that, for efficient recycle, the fuel form should be limited in burnup only by radiation damage to fuel cladding. The recycle technology must recover essentially all actinides. In a fast reactor, not all fission products need to be removed from the recycled fuel, and there is no need to produce pure plutonium. Recovery, recycle, and ultimate consumption of all actinides resolves several waste-disposal concerns. The IFR can be configured to achieve safe passive response to any of the traditional postulated reactor accident initiators, and can be configured for a variety of power output levels. Passive heat removal is achieved by use of a large inventory sodium coolant and a physical configuration that emphasizes natural circulation. An IFR can be designed to consume excess fissile material, to produce a surplus, or to maintain inventory. It appears that commercial designs should be economically competitive with other available alternatives. (author)

Civilian Power Program. Part 1, Summary, Current status of reactor concepts

Energy Technology Data Exchange (ETDEWEB)

Author, Not Given

1959-09-01

This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).

Proposal for a technology-neutral safety approach for new reactor designs

International Nuclear Information System (INIS)

2007-09-01

Many states are considering an expansion of their nuclear power generation programmes. Many of the technologies and concepts are new and innovative. The current design and licensing rules are applicable to mostly large water reactors and there are no accepted rules in place for design, safety assessment and licensing for new innovative nuclear power plants. This TECDOC proposes a (new) safety approach and a methodology to generate technology-neutral (i.e. independent of reactor technology) safety requirements and a 'safe design' for advanced and innovative reactors. The experience gained in decades of design and licensing, combined with the development of risk-based concepts, has provided insights that will form the basis for new safety rules and requirements. Many lessons learned acknowledge the importance of such concepts as safety goals and defence in depth and the benefits of integrating risk insights early in an iterative design process. A new safety approach will incorporate many of the new developments in these concepts. For example, the probabilistic elements of defence in depth will help define the cumulative provisions to compensate for uncertainty and incompleteness of our knowledge of accident initiation and progression. This TECDOC also identifies areas of work, which will require further definition, research and development and guidance on application. This publication is to be used as a guide to developing a new technology-neutral safety approach, and as a guide in the application of methodologies to define the safety requirements for an innovative reactor designs. The method proposes an integration of deterministic and probabilistic considerations with established principles and concepts such as safety goals and defence in depth. The TECDOC recommends that the structure of the new technology-neutral main pillars for the design and licensing of innovative nuclear reactors be developed following a top-down approach to reflect a newer risk-informed and

Improving Safety, Economic, Substantiality, and Security of Nuclear Energy with Canadian Super-Critical Water-cooled Reactor Concept

International Nuclear Information System (INIS)

Hamilton, Holly; Pencer, Jeremy; Yetisir, Metin; Leung, Laurence

2012-01-01

Super-Critical Water-cooled Reactor is one of the six design concepts being developed under the Generation IV International Forum. It is the only concept evolving from the water-cooled reactors and taking advantages of the balance-of-plant design and operation experience of the fossil-power plants. Canada is developing the SCR concept from the well-established pressure-tube reactor technology. The Canadian SCWR maintains modular design approach using relative small fuel channels with the separation of coolant and moderator. It is equipped with an advanced fuel channel design that is capable to transfer decay heat from the fuel to the moderator under the long-term cooling stage. Coupled with the advanced passive-moderator cooling system, cooling of fuel and fuel channel is continuous even without external power or operator intervention. The Canadian SCWR is operating at a pressure of 25 MPa with a core outlet temperature of 625 deg. C. This has led to a drastic increase in thermal efficiency to 48% from 34% of the current fleet of reactors (a 40% rise in relative efficiency). With the high core outlet temperature, a direct thermal cycle has been adopted and has led to simplification in plant design attributing to the cost reduction compared to the current reactor designs. The Canadian SCWR adopts the advanced Thorium fuel cycle to enhance the substantiality, economic, and security. than uranium in the world (estimated to be three times more). This provides the long-term fuel supply. Thorium's price is stable compared to uranium and is consistently lower than uranium. This would maintain the predictability and economic of fuel supply. Thorium itself is a non-fissile material and once irradiated requires special handling. This improves proliferative resistance. The objective of this paper is to highlight these improvements in generating nuclear energy with the Canadian SCWR

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.

Gas reactor international cooperative program interim report: German Pebble Bed Reactor design and technology review

International Nuclear Information System (INIS)

1978-09-01

This report describes and evaluates several gas-cooled reactor plant concepts under development within the Federal Republic of Germany (FRG). The concepts, based upon the use of a proven Pebble Bed Reactor (PBR) fuel element design, include nuclear heat generation for chemical processes and electrical power generation. Processes under consideration for the nuclear process heat plant (PNP) include hydrogasification of coal, steam gasification of coal, combined process, and long-distance chemical heat transportation. The electric plant emphasized in the report is the steam turbine cycle (HTR-K), although the gas turbine cycle (HHT) is also discussed. The study is a detailed description and evaluation of the nuclear portion of the various plants. The general conclusions are that the PBR technology is sound and that the HTR-K and PNP plant concepts appear to be achievable through appropriate continuing development programs, most of which are either under way or planned

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 nuclear power reactor concept for Brazil

International Nuclear Information System (INIS)

Sefidvash, F.

1980-01-01

For the purpose of developing an independent national nuclear technology and effective manner of transferring such a technology, as well as developing a modern reactor, a new nuclear power reactor concept is proposed which is considered as a suitable and viable project for Brazil to support its development and finally construct its prototype as an indigeneous venture. (Author) [pt

Status of advanced light water cooled reactor designs 1996

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-09-01

The present report, which is significantly more comprehensive than the previously one, addresses the rationale and basic motivations that lead to a continuing development of nuclear technology, provides an overview of the world status of current LWRs, describes the present market situations, and identifies desired characteristics for future plants. The report also provides a detailed description of utility requirements that largely govern today`s nuclear development efforts, the situation with regard to enhanced safety objectives, a country wise description of the development activities, and a technical description of the various reactor designs in a consistent format. The reactor designs are presented in two categories: (1) evolutionary concepts that are expected to be commercially available soon; and (2) innovative designs. The report addresses the main technical characteristics of each concept without assessing or evaluating them from a particular point of view (e.g. safety or economics). Additionally, the report identifies basic reference documents that can provide further information for detailed evaluations. The report closes with an outlook on future energy policy developments.

Status of advanced light water cooled reactor designs 1996

International Nuclear Information System (INIS)

1997-09-01

The present report, which is significantly more comprehensive than the previously one, addresses the rationale and basic motivations that lead to a continuing development of nuclear technology, provides an overview of the world status of current LWRs, describes the present market situations, and identifies desired characteristics for future plants. The report also provides a detailed description of utility requirements that largely govern today's nuclear development efforts, the situation with regard to enhanced safety objectives, a country wise description of the development activities, and a technical description of the various reactor designs in a consistent format. The reactor designs are presented in two categories: (1) evolutionary concepts that are expected to be commercially available soon; and (2) innovative designs. The report addresses the main technical characteristics of each concept without assessing or evaluating them from a particular point of view (e.g. safety or economics). Additionally, the report identifies basic reference documents that can provide further information for detailed evaluations. The report closes with an outlook on future energy policy developments

The PRISM concept for a safe, economic and testable liquid metal fast reactor plant

International Nuclear Information System (INIS)

Berglund, R.C.; Salerno, L.N.; Tippets, F.E.

1987-01-01

The PRISM project is underway at General Electric as part of an advanced reactor conceptual design program sponsored by the US Department of Energy. The PRISM concept emphasizes inherent safety, modular construction, and factory fabrication. These features are intended to improve the basis for public acceptance, reduce cost,improve licensability, and reduce the risk of schedule delays and cost increases during construction. A PRISM power plant comprises a number of reactor modules. The relatively small size of the reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features for safe accommodation of accidents. These inherent safety features permit simplification and reduction of conventional safety-related systems in the plant. Testing of a full-size prototype reactor module is planned in the late 1990's to demonstrate these inherent safety characteristics. It is intended that the results of the test be used to obtain certification of the design by the US Nuclear Regulatory Commission preparatory to use of reactor modules built to this standard design in licensed commercial plants

Development of Core Design Model for Small-Sized Research Reactor and Establishment of Infrastructure for Reactor Export

International Nuclear Information System (INIS)

Kim, M. H.; Win, Naing; Lim, J. Y.

2007-02-01

Within 10 years a growing world-wide demand of new research reactor construction is expected because of obsolescence. In Korea, a new research reactor is also required in order to meet domestic demand of utilization. KAERI has been devoted to develop an export-oriented research reactors for these kinds of demand. A next generation research reactor should comply with general requirements for safety, economics, environment-friendliness and non-proliferation as well as high performance requirement of high flux level. A export-tailored reactor should be developed for the demand of developing counties or under-developed countries. A new design concept is to be developed for a long cycle length core which has excellent irradiation facility with high flux

Safety concept of high-temperature reactors based on the experience with AVR and THTR

International Nuclear Information System (INIS)

Wachholz, Winfried; Kroeger, Wolfgang

1990-01-01

In the Federal Republic of Germany a reactor is considered safe if verification has been furnished that the requirements contained in paragraph 7 of the Federal German Atomic Energy Act are met for this reactor: demonstration of sufficient precautions against damage required according to the actual state of the art, and especially compliance with the dose rate limits for normal operation and accidental conditions. These requirements result in a deterministic multi-stage safety concept with specified requirements for the engineered safety systems. In recent years, proposals for enhanced safety of nuclear power reactors or a radical change in safety philosophy have been made. This is characterised by 'inherently safe', 'super safe' and similar slogans. A quantitative definition of these requirements has not yet been established, but it is clear as a common objective that the event of beyond design basis accidents evacuation, relocation, and large scale contamination of ground should not occur. As a consequence of the Chernobyl accident the safety of all the NPPs in Germany has been reviewed. This analysis was completed for the THTR reactor in 1988. The same has been done for AVR reactor. The final evaluation of the HTR specific safety features have been fully confirmed. The HTR concepts under development are based on this experience. The HTR-Modul unit is currently being designed

Description of reactor fuel breeding with three integral concepts

International Nuclear Information System (INIS)

Ott, K.O.; Hanan, N.A.; Maudlin, P.J.; Borg, R.C.

1979-01-01

The time-dependent breeding of fuel in a growing system of breeder reactors can be characterized by the transitory (instantaneous) growth rate, γ(t). The three most important aspects of γ(t) can be expressed by time-independent integral concepts. Two of these concepts are in widespread use. A third integral concept that links the two earlier ones is introduced. The time-dependent growth rate has an asymptotic value, γ/sup infinity/, the equilibrium growth rate, which is the basis for the calculation of the doubling time. The equilibrium growth rate measures the breeding capability and represents a reactor property. Maximum deviation of γ(t) and γ/sup infinity/ generally appears at the initial startup of the reactor, where γ(t = 0) = γ 0 . This deviation is due to the difference between the initial and asymptotic fuel inventory composition. The initial growth rate can be considered a second integral concept; it characterizes the breeding of a particular fuel in a given reactor. Growth rates are logarithmic derivatives of the growing mass of fuel in breeder reactors, especially γ/sup infinity/, which describes the asymptotic growth by exp(γ/sup infinity/t). There is, however, a variation in the fuel-mass factor in front of this exponential function during the transition from γ 0 to γ/sup infinity/. It is shown that this variation of the fuel mass during transitioncan be described by a third integral concept, termed the breeding bonus, b. The breeding bonus measures the quality of a fuel for its use in a given reactor in terms of its impact on the magnitude of the asymptotically growing fuel mass. The calculation of γ 0 and γ/sup infinity/ is facilitated by use of the critical mass (CM) worths and the breeding worth factors, respectively

ELMO Bumpy Torus Reactor and power plant: conceptual design study

International Nuclear Information System (INIS)

Bathke, C.G.; Dudziak, D.J.; Krakowski, R.A.

1981-08-01

A complete power plant design of a 1200-MWe ELMO Bumpy Torus Reactor (EBTR) is presented. An emphasis is placed on those features that are unique to the EBT confinement concept, with subsystems and balance-of-plant items that are more generic to magnetic fusion being adapted from past, more extensive tokamak reactor designs. Similar to the latter tokamak studies, this conceptual EBTR design also emphasizes the use of conventional or near state-of-the-art engineering technology and materials. An emphasis is also placed on system accessibility, reliability, and maintainability, as these crucial and desirable characteristics relate to the unique high-aspect-ratio configuration of EBTs. Equal and strong emphasis is given to physics, engineering/technology, and costing/economics components of this design effort. Parametric optimizations and sensitivity studies, using cost-of-electricity as an object function, are reported. Based on these results, the direction for future improvement on an already attractive reactor design is identified

Recent developments in engineering and technology concepts for prospective tokamak fusion reactors

International Nuclear Information System (INIS)

Ford, G.W.K.

1987-01-01

The tokamak has become the most developed magnetic fusion system and it appears likely that break-even and possibly ignition will first be demonstrated in existing machines of this type. Yet larger tokamaks could also demonstrate the essential technologies for the production of useful power. World-wide, well over a hundred tritium-breeder/heat-removal blanket concepts have been devised and preliminary engineering design studies undertaken, but the effort deployed on breeding and power recovery systems has been very small compared with that assigned to plasma research and development. The European Communities' NET (Next European Torus) project may offer an opportunity to redress this imbalance. The NET pre-design stage now in progress for some three years has selected many of the best features of plasma and nuclear design from the world's total efforts in these fields, and the NET concept is described in this paper as exemplifying where magnetic fusion power reactor technology stands today. It is concluded that although there are numerous more advanced types of magnetic confinement fusion reactor at early stages of their physics development, the tokamak offers the best opportunity for the early demonstration of fusion power

Major NSSS design features of the Korean next generation reactor

International Nuclear Information System (INIS)

Kim, Insk; Kim, Dong-Su

1999-01-01

In order to meet national needs for increasing electric power generation in the Republic of Korea in the 2000s, the Korean nuclear development group (KNDG) is developing a standardized evolutionary advanced light water reactor (ALWR), the Korean Next Generation Reactor (KNGR). It is an advanced version of the successful Korean Standard Nuclear Power Plant (KSNP) design, which meets utility needs for safety enhancement, performance improvement and ease of operation and maintenance. The KNGR design starts fro the proven design concept of the currently operating KSNPs with uprated power and advanced design features required by the utility. The KNGR design is currently in the final stage of the basic design, and the paper describes the major nuclear steam supply system (NSSS) design features of the KNGR together with introduction of the KNGR development program. (author)

Development of small and medium integral reactor. ctor Development of fluid system design for small and medium integral reactor

International Nuclear Information System (INIS)

Lee, D. J.; Chang, M. H.; Kim, K. K.; Kim, J. P.; Yoon, J. H.; Lee, Y. J.; Park, C. T.; Bae, Y. Y.; Kang, D. J.; Lee, K. H.; Lee, J.; Kim, H. Y.; Cho, B. H.; Seo, J. K.; Kang, K. S.; Kang, H. O.

1997-07-01

The purpose of this study is to develop system design technology of integral reactor, as a new design concept of small and medium reactor having enhanced safety and economy, and to have a design assessment / verification technology through basic thermal hydraulic experiments. This report describes of the following: 1) basic requirement for the integral reactor system design 2) Conceptual design of primary and secondary circuits of NSSS, emergency core cooling system, passive residual heat removal system, severe accident mitigation cooling system, passive residual heat removal system, severe accident mitigation system and other auxiliary system 3) Requirements and test program for the basic thermal hydraulic experiments including, CHF test for hexagonal fuel assembly, flow instability for once-through steam generator, core flow distribution test and verification test for non-condensable gas model in RELAP-5 code. The results of this study can be utilized for using as the foundation technology of in the next basic design phase and design technology for future advanced reactors. (author). 30 refs.,24 tabs., 56 figs

Development of small and medium integral reactor. ctor Development of fluid system design for small and medium integral reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, D. J.; Chang, M. H.; Kim, K. K.; Kim, J. P.; Yoon, J. H.; Lee, Y. J.; Park, C. T.; Bae, Y. Y.; Kang, D. J.; Lee, K. H.; Lee, J.; Kim, H. Y.; Cho, B. H.; Seo, J. K.; Kang, K. S.; Kang, H. O.

1997-07-01

The purpose of this study is to develop system design technology of integral reactor, as a new design concept of small and medium reactor having enhanced safety and economy, and to have a design assessment / verification technology through basic thermal hydraulic experiments. This report describes of the following: (1) basic requirement for the integral reactor system design (2) Conceptual design of primary and secondary circuits of NSSS, emergency core cooling system, passive residual heat removal system, severe accident mitigation cooling system, passive residual heat removal system, severe accident mitigation system and other auxiliary system (3) Requirements and test program for the basic thermal hydraulic experiments including, CHF test for hexagonal fuel assembly, flow instability for once-through steam generator, core flow distribution test and verification test for non-condensable gas model in RELAP-5 code. The results of this study can be utilized for using as the foundation technology of in the next basic design phase and design technology for future advanced reactors. (author). 30 refs.,24 tabs., 56 figs.

Changing concepts of geologic structure and the problem of siting nuclear reactors: examples from Washington State

International Nuclear Information System (INIS)

Tabor, R.W.

1986-01-01

The conflict between regulation and healthy evolution of geological science has contributed to the difficulties of siting nuclear reactors. On the Columbia Plateau in Washington, but for conservative design of the Hanford reactor facility, the recognition of the little-understood Olympic-Wallowa lineament as a major, possibly still active structural alignment might have jeopardized the acceptability of the site for nuclear reactors. On the Olympic Peninsula, evolving concepts of compressive structures and their possible recent activity and the current recognition of a subducting Juan de Fuca plate and its potential for generating great earthquakes - both concepts little-considered during initial site selection - may delay final acceptance of the Satsop site. Conflicts of this sort are inevitable but can be accommodated if they are anticipated in the reactor-licensing process. More important, society should be increasing its store of geologic knowledge now, during the current recess in nuclear reactor siting

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2011-07-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

International Nuclear Information System (INIS)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

2011-01-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

State of the art of the fluidized bed nuclear reactor concept

International Nuclear Information System (INIS)

Sefidvash, F.; Vilhena, M.T.M.B. de; Streck, E.; Borges, V.; Johansson, M.

1987-01-01

A small and simple nuclear reactors with inherent safety using the fluidized bed concept is under research and study. In this paper a brief study neutronics and thermal hydraulics of this reactor concept is presented. (Author) [pt

Multi-physics design and analyses of long life reactors for lunar outposts

Science.gov (United States)

Schriener, Timothy M.

Future human exploration of the solar system is likely to include establishing permanent outposts on the surface of the Moon. These outposts will require reliable sources of electrical power in the range of 10's to 100's of kWe to support exploration and resource utilization activities. This need is best met using nuclear reactor power systems which can operate steadily throughout the long ˜27.3 day lunar rotational period, irrespective of location. Nuclear power systems can potentially open up the entire lunar surface for future exploration and development. Desirable features of nuclear power systems for the lunar surface include passive operation, the avoidance of single point failures in reactor cooling and the integrated power system, moderate operating temperatures to enable the use of conventional materials with proven irradiation experience, utilization of the lunar regolith for radiation shielding and as a supplemental neutron reflector, and safe post-operation decay heat removal and storage for potential retrieval. In addition, it is desirable for the reactor to have a long operational life. Only a limited number of space nuclear reactor concepts have previously been developed for the lunar environment, and these designs possess only a few of these desirable design and operation features. The objective of this research is therefore to perform design and analyses of long operational life lunar reactors and power systems which incorporate the desirable features listed above. A long reactor operational life could be achieved either by increasing the amount of highly enriched uranium (HEU) fuel in the core or by improving the neutron economy in the reactor through reducing neutron leakage and parasitic absorption. The amount of fuel in surface power reactors is constrained by the launch safety requirements. These include ensuring that the bare reactor core remains safely subcritical when submerged in water or wet sand and flooded with seawater in the unlikely