Status and some safety philosophies of the China advanced research reactor CARR

International Nuclear Information System (INIS)

Luzheng Yuan

2001-01-01

The existing two research reactors, HWRR (heavy water research reactor) and SPR (swimming pool reactor), have been operated by China Institute of Atomic Energy (CIAE) since, respectively, 1958 and 1964, and are both in extending service and facing the aging problem. It is expected that they will be out of service successively in the beginning decade of the 21 st century. A new, high performance and multipurpose research reactor called China advanced research reactor (CARR) will replace these two reactors. This new reactor adopts the concept of inverse neutron trap compact core structure with light water as coolant and heavy water as the outer reflector. Its design goal is as follows: under the nuclear power of 60MW, the maximum unperturbed thermal neutron flux in peripheral D 2 O reflector not less than 8 x 10 14 n/cm 2 . s while in central experimental channel, if the central cell to be replaced by an experimental channel, the corresponding value not less than 1 x 10 15 n/cm 2 . s. The main applications for this research reactor will cover RI production, neutron scattering experiments, NAA and its applications, neutron photography, NTD for monocrystaline silicon and applications on reactor engineering technology. By the end of 1999, the preliminary design of CARR was completed, then the draft of preliminary safety analysis report (PSAR) was submitted to the relevant authority at the end of 2000 for being reviewed. Now, the CARR project has entered the detail design phase and safety reviewing procedure for obtaining the construction permit from the relevant licensing authority. This paper will only briefly introduce some aspects of safety philosophy of CARR design and PSAR. (orig.)

Research and development on the application of advanced control technologies to advanced nuclear reactor systems: A US national perspective

International Nuclear Information System (INIS)

White, J.D.; Monson, L.R.; Carrol, D.G.; Dayal, Y.

1989-01-01

Control system designs for nuclear power plants are becoming more advanced through the use of digital technology and automation. This evolution is taking place because of: (1) the limitations in analog based control system performance and maintenance and availability and (2) the promise of significant improvement in plant operation and availability due to advances in digital and other control technologies. Digital retrofits of control systems in US nuclear plants are occurring now. Designs of control and protection systems for advanced LWRs are based on digital technology. The use of small inexpensive, fast, large-capacity computers in these designs is the first step of an evolutionary process described in this paper. Under the sponsorship of the US Department of Energy (DOE), Oak Ridge National Laboratory, Argonne National Laboratory, GE Nuclear Energy and several universities are performing research and development in the application of advances in control theory, software engineering, advanced computer architectures, artificial intelligence, and man-machine interface analysis to control system design. The target plant concept for the work described in this paper is the Power Reactor Inherently Safe Module reactor (PRISM), an advanced modular liquid metal reactor concept. This and other reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. 18 refs., 5 figs

Operation and utilization of low power research reactor critical facility for Advanced Heavy Water Reactor (AHWR)

International Nuclear Information System (INIS)

De, S.K.; Karhadkar, C.G.

2017-01-01

An Advanced Heavy Water Reactor (AHWR) has been designed and developed for maximum power generation from thorium considering large reserves of thorium. The design envisages using 54 pin MOX cluster with different enrichment of "2"3"3U and Pu in Thoria fuel pins. Theoretical models developed to neutron transport and the geometrical details of the reactor including all reactivity devices involve approximations in modelling, resulting in uncertainties. With a view to minimize these uncertainties, a low power research reactor Critical Facility was built in which cold clean fuel can be arranged in a desired and precise geometry. Different experiments conducted in this facility greatly contribute to understand and validate the physics design parameters

Research reactors - an overview

International Nuclear Information System (INIS)

West, C.D.

1997-01-01

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

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

Progress in Methodologies for the Assessment of Passive Safety System Reliability in Advanced Reactors. Results from the Coordinated Research Project on Development of Advanced Methodologies for the Assessment of Passive Safety Systems Performance in Advanced Reactors

International Nuclear Information System (INIS)

2014-09-01

Strong reliance on inherent and passive design features has become a hallmark of many advanced reactor designs, including several evolutionary designs and nearly all advanced small and medium sized reactor (SMR) designs. Advanced nuclear reactor designs incorporate several passive systems in addition to active ones — not only to enhance the operational safety of the reactors but also to eliminate the possibility of serious accidents. Accordingly, the assessment of the reliability of passive safety systems is a crucial issue to be resolved before their extensive use in future nuclear power plants. Several physical parameters affect the performance of a passive safety system, and their values at the time of operation are unknown a priori. The functions of passive systems are based on basic physical laws and thermodynamic principals, and they may not experience the same kind of failures as active systems. Hence, consistent efforts are required to qualify the reliability of passive systems. To support the development of advanced nuclear reactor designs with passive systems, investigations into their reliability using various methodologies are being conducted in several Member States with advanced reactor development programmes. These efforts include reliability methods for passive systems by the French Atomic Energy and Alternative Energies Commission, reliability evaluation of passive safety system by the University of Pisa, Italy, and assessment of passive system reliability by the Bhabha Atomic Research Centre, India. These different approaches seem to demonstrate a consensus on some aspects. However, the developers of the approaches have been unable to agree on the definition of reliability in a passive system. Based on these developments and in order to foster collaboration, the IAEA initiated the Coordinated Research Project (CRP) on Development of Advanced Methodologies for the Assessment of Passive Safety Systems Performance in Advanced Reactors in 2008. The

Advanced fuel in the Budapest research reactor

International Nuclear Information System (INIS)

Hargitai, T.; Vidovsky, I.

1997-01-01

The Budapest Research Reactor, the first nuclear facility of Hungary, started to operate in 1959. The main goal of the reactor is to serve neutron research, but applications as neutron radiography, radioisotope production, pressure vessel surveillance test, etc. are important as well. The Budapest Research Reactor is a tank type reactor, moderated and cooled by light water. After a reconstruction and upgrading in 1967 the VVR-SM type fuel elements were used in it. These fuel elements provided a thermal power of 5 MW in the period 1967-1986 and 10 MW after the reconstruction from 1992. In the late eighties the Russian vendor changed the fuel elements slightly, i.e. the main parameters of the fuel remained unchanged, however a higher uranium content was reached. This new fuel is called VVR-M2. The geometry of VVR-SM and VVR-M2 are identical, allowing the use to load old and new fuel assemblies together to the active core. The first new type fuel assemblies were loaded to the Budapest Research Reactor in 1996. The present paper describes the operational experience with the new type of fuel elements in Hungary. (author)

Prospects for the development of advanced reactors. [Advanced Reactors

Energy Technology Data Exchange (ETDEWEB)

Semenov, B. A.; Kupitz, J.; Cleveland, J. [International Atomic Energy Agency Vienna (Austria). Dept. of Nuclear Energy and Safety

1992-01-01

Energy supply is an important prerequisite for further socio-economic development, especially in developing countries where the per capita energy use is only a very small fraction of that in industrialized countries. Nuclear energy is an essentially unlimited energy resource with the potential to provide this energy in the form of electricity, district heat and process heat under environmentally acceptable conditions. However, this potential will be realized only if nuclear power plants can meet the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide a tremendous amount of experience has been accumulated during development, licensing, construction and operation of nuclear power reactors. The experience forms a sound basis for further improvements. Nuclear programmes in many countries are addressing the development of advanced reactors which are intended to have better economics, higher reliability and improved safety in order to overcome the current concerns of nuclear power. Advanced reactors now being developed could help to meet the demand for new plants in developed and developing countries, not only for electricity generation, but also for district heating, desalination and for process heat. The IAEA, as the only global international governmental organization dealing with nuclear power, promotes international information exchange and international co-operation between all countries with their own advanced nuclear power programmes and offers assistance to countries with an interest in exploratory or research programmes.

Advanced reactor development

International Nuclear Information System (INIS)

Till, C.E.

1989-01-01

Consideration is given to what the aims of advanced reactor development have to be, if a new generation of nuclear power is really to play an important role in man's energy generation activities in a fragile environment. The background given briefly covers present atmospheric evidence, the current situation in nuclear power, how reactors work and what can go wrong with them, and the present magnitudes of world energy generation. The central part of the paper describes what is currently being done in advanced reactor development and what can be expected from various systems and various elements of it. A vigorous case is made that three elements must be present in any advanced reactor development: (1) breeding; (2) passive safety; and (3) shorter-live nuclear waste. All three are possible. In the right advanced reactor systems the ways of achieving them are known. But R and D is necessary. That is the central argument made in the paper. Not advanced reactor prototype construction at this point, but R and D itself. (author)

Analysis of soil-structure interaction and floor response spectrum of reactor building for China advanced research reactor

International Nuclear Information System (INIS)

Rong Feng; Wang Jiachun; He Shuyan

2006-01-01

Analysis of Soil-Structure Interaction (SSI) and calculation of Floor Response Spectrum (FRS) is substantial for anti-seismic design for China Advanced Research Reactor (CARR) project. The article uses direct method to analyze the seismic reaction of the reactor building in considering soil-structure interaction by establishing two-dimensional soil-structure co-acting model for analyzing and inputting of seismic waves from three directions respectively. The seismic response and floor response spectrum of foundation and floors of the building under different cases have been calculated. (authors)

Plant maintenance and advanced reactors, 2006

Energy Technology Data Exchange (ETDEWEB)

Agnihotri, Newal (ed.)

2006-09-15

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Advanced plants to meet rising expectations, by John Cleveland, International Atomic Energy Agency, Vienna; A flexible and economic small reactor, by Mario D. Carelli and Bojan Petrovic, Westinghouse Electric Company; A simple and passively safe reactor, by Yury N. Kuznetsov, Research and Development Institute of Power Engineering (NIKIET), Russia; Gas-cooled reactors, by Jeffrey S. Merrifield, U.S. Nuclear Regulatory Commission; ISI project managment in the PRC, by Chen Chanbing, RINPO, China; and, Fort Calhoun refurbishment, by Sudesh Cambhir, Omaha Public Power District.

Development of a Fissile Materials Irradiation Capability for Advanced Fuel Testing at the MIT Research Reactor

International Nuclear Information System (INIS)

Hu Linwen; Bernard, John A.; Hejzlar, Pavel; Kohse, Gordon

2005-01-01

A fissile materials irradiation capability has been developed at the Massachusetts Institute of Technology (MIT) Research Reactor (MITR) to support nuclear engineering studies in the area of advanced fuels. The focus of the expected research is to investigate the basic properties of advanced nuclear fuels using small aggregates of fissile material. As such, this program is intended to complement the ongoing fuel evaluation programs at test reactors. Candidates for study at the MITR include vibration-packed annular fuel for light water reactors and microparticle fuels for high-temperature gas reactors. Technical considerations that pertain to the design of the MITR facility are enumerated including those specified by 10 CFR 50 concerning the definition of a research reactor and those contained in a separate license amendment that was issued by the U.S. Nuclear Regulatory Commission to MIT for these types of experiments. The former includes limits on the cross-sectional area of the experiment, the physical form of the irradiated material, and the removal of heat. The latter addresses experiment reactivity worth, thermal-hydraulic considerations, avoidance of fission product release, and experiment specific temperature scrams

Development of Regulatory Technical Requirements for the Advanced Integral Type Research Reactor

International Nuclear Information System (INIS)

Jo, Jong Chull; Yune, Young Gill; Kim, Woong Sik; Kim, Hho Jung

2004-01-01

This paper presents the current status of the study on the development of regulatory technical requirements for the licensing review of an advanced integral type research reactor of which the license application is expected in a few years. According to the Atomic Energy Act of Korea, both research and education reactors are subject to the technical requirements for power reactors in the licensing review. But, some of the requirements may not be applicable or insufficient for the licensing reviews of reactors with unique design features. Thus it is necessary to identify which review topics or areas can not be addressed by the existing requirements and to develop the required ones newly or supplement appropriately. Through the study performed so far, it has been identified that the following requirements need to be developed newly for the licensing review of SMART-P: the use of proven technology, the interfacial facility, the non-safety systems, and the metallic fuels. The approach and basis for the development of each of the requirements are discussed. (authors)

CSAU (code scaling, applicability and uncertainty), a tool to prioritize advanced reactor research

International Nuclear Information System (INIS)

Wilson, G.E.; Boyack, B.E.

1990-01-01

Best Estimate computer codes have been accepted by the US Nuclear Regulatory Commission as an optional tool for performing safety analysis related to the licensing and regulation of current nuclear reactors producing commercial electrical power, providing their uncertainty is quantified. In support of this policy change, the NRC and its contractors and consultants have developed and demonstrated an uncertainty quantification methodology called CSAU. At the process level, the method is generic to any application which relies on best estimate computer code simulations to determine safe operating margins. The primary use of the CSAU methodology is to quantify safety margins for existing designs; however, the methodology can also serve an equally important role in advanced reactor research for plants not yet built. Applied early, during the period when alternate designs are being evaluated, the methodology can identify the relative importance of the sources of uncertainty in the knowledge of each plant behavior and, thereby, help prioritize the research needed to bring the new designs to fruition. This paper describes the CSAU methodology, at the generic process level, and provides the general principles whereby it may be applied to evaluations of advanced reactor designs. 9 refs., 1 fig., 1 tab

Research and Development of Protection OPC server for China advanced research reactor digital monitoring system

International Nuclear Information System (INIS)

Jia Yuwen; Xu Qiguo

2012-01-01

OPC server was developed as I/O driver to communicate the digital monitoring system of China Advanced Research Reactor iFIX and protection system. The framework and working principle of the OPC server were researched, and an effective method was developed to resolve the special communication protocol. After commissioning and testing, the results show that this method is reliable and stable, makes the system easy to configure, and can reduce the complexity of the system. (authors)

Design requirement for electrical system of an advanced research reactor

International Nuclear Information System (INIS)

Jung, Hoan Sung; Kim, H. K.; Kim, Y. K.; Wu, J. S.; Ryu, J. S.

2004-12-01

An advanced research reactor is being designed since 2002 and the conceptual design has been completed this year for the several types of core. Also the fuel was designed for the potential cores. But the process system, the I and C system, and the electrical system design are under pre-conceptual stage. The conceptual design for those systems will be developed in the next year. Design requirements for the electrical system set up to develop conceptual design. The same goals as reactor design - enhance safety, reliability, economy, were applied for the development of the requirements. Also the experience of HANARO design and operation was based on. The design requirements for the power distribution, standby power supply, and raceway system will be used for the conceptual design of electrical system

Design requirement for electrical system of an advanced research reactor

Energy Technology Data Exchange (ETDEWEB)

Jung, Hoan Sung; Kim, H. K.; Kim, Y. K.; Wu, J. S.; Ryu, J. S

2004-12-01

An advanced research reactor is being designed since 2002 and the conceptual design has been completed this year for the several types of core. Also the fuel was designed for the potential cores. But the process system, the I and C system, and the electrical system design are under pre-conceptual stage. The conceptual design for those systems will be developed in the next year. Design requirements for the electrical system set up to develop conceptual design. The same goals as reactor design - enhance safety, reliability, economy, were applied for the development of the requirements. Also the experience of HANARO design and operation was based on. The design requirements for the power distribution, standby power supply, and raceway system will be used for the conceptual design of electrical system.

Numerical simulation of flow field in the China advanced research reactor flow-guide tank

International Nuclear Information System (INIS)

Xu Changjiang

2002-01-01

The flow-guide tank in China advanced research reactor (CARR) acts as a reactor inlet coolant distributor and play an important role in reducing the flow-induced vibration of the internal components of the reactor core. Numerical simulations of the flow field in the flow-guide tank under different conceptual designing configurations are carried out using the PHOENICS3.2. It is seen that the inlet coolant is well distributed circumferentially into the flow-guide tank with the inlet buffer plate and the flow distributor barrel. The maximum cross-flow velocity within the flow-guide tank is reduced significantly, and the reduction of flow-induced vibration of reactor internals is expected

Materials for advanced water cooled reactors

International Nuclear Information System (INIS)

1992-09-01

The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

Experimental facilities for gas-cooled reactor safety studies. Task group on Advanced Reactor Experimental Facilities (TAREF)

International Nuclear Information System (INIS)

2009-01-01

In 2007, the NEA Committee on the Safety of Nuclear Installations (CSNI) completed a study on Nuclear Safety Research in OECD Countries: Support Facilities for Existing and Advanced Reactors (SFEAR) which focused on facilities suitable for current and advanced water reactor systems. In a subsequent collective opinion on the subject, the CSNI recommended to conduct a similar exercise for Generation IV reactor designs, aiming to develop a strategy for ' better preparing the CSNI to play a role in the planned extension of safety research beyond the needs set by current operating reactors'. In that context, the CSNI established the Task Group on Advanced Reactor Experimental Facilities (TAREF) in 2008 with the objective of providing an overview of facilities suitable for performing safety research relevant to gas-cooled reactors and sodium fast reactors. This report addresses gas-cooled reactors; a similar report covering sodium fast reactors is under preparation. The findings of the TAREF are expected to trigger internationally funded CSNI projects on relevant safety issues at the key facilities identified. Such CSNI-sponsored projects constitute a means for efficiently obtaining the necessary data through internationally co-ordinated research. This report provides an overview of experimental facilities that can be used to carry out nuclear safety research for gas-cooled reactors and identifies priorities for organizing international co-operative programmes at selected facilities. The information has been collected and analysed by a Task Group on Advanced Reactor Experimental Facilities (TAREF) as part of an ongoing initiative of the NEA Committee on the Safety of Nuclear Installations (CSNI) which aims to define and to implement a strategy for the efficient utilisation of facilities and resources for Generation IV reactor systems. (author)

LBB application in the US operating and advanced reactors

Energy Technology Data Exchange (ETDEWEB)

Wichman, K.; Tsao, J.; Mayfield, M.

1997-04-01

The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

Development of essential system technologies for advanced reactor

International Nuclear Information System (INIS)

Bae, Y. Y.; Hwang, Y. D.; Cho, B. H. and others

1999-03-01

Basic design of SMART adopts the new advanced technologies which were not applied in the existing 1000MWe PWR. However, the R and D experience on these advanced essential technologies is lacking in domestic nuclear industry. Recently, a research on these advanced technologies has been performed as a part of the mid-and-long term nuclear R and D program, but the research was limited only for the small scale fundamental study. The research on these essential technologies such as helically coiled tube steam generator, self pressurizer, core cooling by natural circulation required for the development of integral reactor SMART have not been conducted in full scale. This project, therefore, was performed for the development of analysis models and methodologies, system analysis and thermal hydraulic experiments on the essential technologies to be applied to the 300MWe capacity of integral reactor SMART and the advanced passive reactor expected to be developed in near future with the emphasis on experimental investigation. (author)

Recent advances in severe accident technology - direct containment heating in advanced light water reactors

International Nuclear Information System (INIS)

Fontana, M.H.

1993-01-01

The issues affecting high-pressure melt ejection (HPME) and the consequential containment pressurization from direct containment heating (DCH), as they affect advanced light water reactors (ALWRs), specifically advanced pressurized water reactors (APWRs), were reviewed by the U.S. Department of Energy Advanced Reactor Severe Accident Program (ARSAP). Recommendations from ARSAP regarding the design of APWRs to minimize DCH are embodied within the Electric Power Research Institute ALWR Utility Requirements Document, which specifies (a) a large, strong containment; (b) an in-containment refueling water storage tank; (c) a reactor cavity configuration that minimizes energy transport to the containment atmosphere; and (d) a reactor coolant system depressurization system. Experimental and analytical efforts, which have focused on current-generation plants, and analyses for APWRs were reviewed. Although DCH is a subject of continuous research and considerable uncertainties remain, it is the judgment of the ARSAP that reactors complying with the recommended design requirements would have a low probability of early containment failure due to HPME and DCH

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

Recent activities of the international Group on Research Reactors (IGORR) and of the Advanced Neutron Source (ANS)

International Nuclear Information System (INIS)

West, C.D.

1992-01-01

The International Group on Research Reactors (IGORR) was formed in 1990 to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. The Advanced Neutron Source Project expects to complete conceptual design in mid-1992. In the present design concept, the neutron source is a heavy-water-cooled, moderated, and reflected reactor of about 350 MW(f) power. (author)

Thermohydraulic relationships for advanced water cooled reactors

International Nuclear Information System (INIS)

2001-04-01

This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

Thermohydraulic relationships for advanced water cooled reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-04-01

This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

DOE/NE University Program in robotics for advanced reactors research

International Nuclear Information System (INIS)

Trivedi, M.M.

1990-01-01

The document presents the bimonthly progress reports published during 1990 regarding the US Department of Energy/NE-sponsored research at the University of Tennessee Knoxville under the DOE Robitics for Advanced Reactors Research Grant. Significant accomplishments are noted in the following areas: development of edge-segment based stereo matching algorithm; vision system integration in the CESAR laboratory; evaluation of algorithms for surface characterization from range data; comparative study of data fusion techniques; development of architectural framework, software, and graphics environment for sensor-based robots; algorithms for acquiring tactile images from planer surfaces; investigations in geometric model-based robotic manipulation; investigations of non-deterministic approaches to sensor fusion; and evaluation of sensor calibration techniques. (MB)

Indian advanced nuclear reactors

International Nuclear Information System (INIS)

Saha, D.; Sinha, R.K.

2005-01-01

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

Cladding and Duct Materials for Advanced Nuclear Recycle Reactors

International Nuclear Information System (INIS)

Allen, Todd R.; Busby, J. T.; Klueh, R. L.; Maloy, Stuart A.; Toloczko, Mychailo B.

2008-01-01

This is a review article that provides an overview of the reactor core structural materials and clad and duct needs for the GNEP advanced burner reactor design. A short history of previous research on structural materials for irradiation environments is provided. There is also a section describing some advanced materials that may be candidate materials for various reactor core structures

Safety features and research needs of westinghouse advanced reactors

International Nuclear Information System (INIS)

Carelli, M.D.; Winters, J.W.; Cummins, W.E.; Bruschi, H.J.

2002-01-01

The three Westinghouse advanced reactors - AP600, AP1000 and IRIS - are at different levels of readiness. AP600 has received a Design Certification, its larger size version AP1000 is currently in the design certification process and IRIS has just completed its conceptual design and will initiate soon a licensing pre-application. The safety features of the passive designs AP600/AP1000 are presented, followed by the features of the more revolutionary IRIS, a small size modular integral reactor. A discussion of the IRIS safety by design approach is given. The AP600/AP1000 design certification is backed by completed testing and development which is summarized, together with a research program currently in progress which will extend AP600 severe accident test data to AP1000 conditions. While IRIS will of course rely on applicable AP600/1000 data, a very extensive testing campaign is being planned to address all the unique aspects of its design. Finally, IRIS plans to use a risk-informed approach in its licensing process. (authors)

The state of art report on advanced reactor development

International Nuclear Information System (INIS)

Kim, Keung Koo; Noh, J. M.; Hwang, D. H. and others

1999-07-01

Recently, researches on the advanced power reactors are being performed actively, that maximize the economics and enhance the reactor safety by introducing the inherent safety characteristics and passive safety features. In the development of advanced reactor technology, we developed the inherent core design technologies which can form a foundation of indigenous technologies to provide the basic technology for the core design of the domestic advanced reactor. In this report, we examined the neutronics design technologies and core thermal hydraulics design technologies for advanced reactors performed all over the world. Major efforts are focussed on the soluble boron free core design technology and high conversion core design technology. In addition to these, new conceptual core, such as a supercritical core, design technology development was also reviewed. The characteristics of critical heat flux have been investigated for non-square lattice rod bundles, such as triangular lattice and wire wrap lattice. Based on the status of advanced reactor development, the soluble boron free and hexagonal lattice core design technologies are elementary technology for the domestic advanced reactor core. These elementary core technologies would enhance the reactor safety and improve the economics. (author). 71 refs., 31 tabs., 74 figs

Revision of construction plan for advanced thermal demonstration reactor

International Nuclear Information System (INIS)

1996-01-01

The Federation of Electric Power Companies demanded the revision of the construction plan for the advanced thermal demonstration reactor, which is included in the 'Long term plan on the research, development and utilization of atomic energy' decided by the Atomic Energy Commission in 1994, for economical reason. The Atomic Energy Commission carried out the deliberation on this demand. It was found that the cost of construction increases to 580 billion yen, and the cost of electric power generation increases three times as high as that of LWRs. The role as the reactor that utilizes MOX fuel can be substituted by LWRs. The relation of trust with the local town must be considered. In view of these circumstances, it is judged that the stoppage of the construction plan is appropriate. It is necessary to investigate the substitute plan for the stoppage, and the viewpoints of investigating the substitute plan, the examination of the advanced BWR with all MOX fuel core and the method of advancing its construction are considered. On the research and development related to advanced thermal reactors, the research and development contributing to the advance of nuclear fuel recycling are advanced, and the prototype reactor 'Fugen' is utilized. (K.I.)

Research reactor utilization. Summary reports of three study group meetings: Irradiation techniques at research reactors, held in Istanbul 15-19 November 1965; Research reactor operation and maintenance problems, held in Caracas 6-10 December 1965; and Research reactor utilization in the Far East, held in Lucas Heights 28 February - 4 March 1966

International Nuclear Information System (INIS)

1967-01-01

The three sections of this book, which are summary reports of three Study Group meetings of the IAEA: Irradiation techniques at research reactors, Istanbul, 15-19 November 1965; Research reactor operation and maintenance problems, Caracas, 6-10 December 1965; and Research reactor utilization in the Far East, Lucas Heights, Australia, 28 February - 4 March 1966. These meetings were the latest in a series designed to promote efficient utilization of research reactors, to disseminate information on advances in techniques, to discuss common problems in reactor operations, and to outline some advanced areas of reactor-based research. (author)

Nuclear Capacity Building through Research Reactors

International Nuclear Information System (INIS)

2017-01-01

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

NRC review of Electric Power Research Institute's Advanced Light Reactor Utility Requirements Document - Program summary, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 1 of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Program Summary,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER provides a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

Advanced Demonstration and Test Reactor Options Study

Energy Technology Data Exchange (ETDEWEB)

Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Gehin, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (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); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States); Qualls, A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Croson, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-01-01

Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy�

Advanced Demonstration and Test Reactor Options Study

International Nuclear Information System (INIS)

Petti, David Andrew; Hill, R.; Gehin, J.; Gougar, Hans David; Strydom, Gerhard; Heidet, F.; Kinsey, J.; Grandy, Christopher; Qualls, A.; Brown, Nicholas; Powers, J.; Hoffman, E.; Croson, D.

2017-01-01

Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power's share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy's (DOE's) broader commitment to pursuing an 'all of the above' clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate 'advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy'. Advanced reactors are

Advanced robotic remote handling system for reactor dismantlement

International Nuclear Information System (INIS)

Shinohara, Yoshikuni; Usui, Hozumi; Fujii, Yoshio

1991-01-01

An advanced robotic remote handling system equipped with a multi-functional amphibious manipulator has been developed and used to dismantle a portion of radioactive reactor internals of an experimental boiling water reactor in the program of reactor decommissioning technology development carried out by the Japan Atomic Energy Research Institute. (author)

Implementation of digital control and protection systems of China advanced research reactor

International Nuclear Information System (INIS)

Zeng Hai; Jin Huajin; Xu Qiguo; Zhang Mingkui

2005-01-01

China Advanced Research Reactor (CARR), a reactor of the 21st century with high performance is being constructed in China. The requirements of reliability and stability on the control and protection (c and p) system are the main points raised. Especially, with the development of digital technology, the c and p system of CARR is demanded to match the trend of digitization in the field of reactor control. The c and p system, including reactor protection system, reactor monitoring and control system, reactor power regulating system, and the mitigation system for ATWS (Anticipate Transient Without Scram), adopts digital technology, and the digital display screen will replace the analog panels in the main control room. The c and p system of CARR adopts redundant technology with 2 or 3 redundant channels to improve the system reliability. The 10/100 Mbps self-adaptive redundant optic fiber industry Ethernet ring network is used to interlink operator workstations, supervisor workstation, and I/O control stations. Commercial grade equipment with mature experience in industrial application are applied to the c and p system of CARR, which have high reliability, good interchangeability, and is easily purchased, the software-developing tools fully match the international industry standards. The realization of digital c and p system of CARR will promote the progress of digital control technology for reactors in China, and certainly become a technical basic platform for developing informational and intelligent reactors in China. (authors)

The role of the IAEA in advanced technologies for water-cooled reactors

International Nuclear Information System (INIS)

Cleveland, J.

1996-01-01

The role of the IAEA in advanced technologies for water-cooled reactors is described, including the following issues: international collaboration ways through international working group activities; IAEA coordinated research programmes; cooperative research in advanced water-cooled reactor technology

A wall-crawling robot for reactor vessel inspection in advanced reactors

International Nuclear Information System (INIS)

Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

1994-01-01

A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected. (author)

A wall-crawling robot for reactor vessel inspection in advanced reactors

International Nuclear Information System (INIS)

Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

1994-01-01

A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected

Advances in reactor physics education: Visualization of reactor parameters

International Nuclear Information System (INIS)

Snoj, L.; Kromar, M.; Zerovnik, G.

2012-01-01

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

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

Investigation of Classification and Design Requirements for Digital Software for Advanced Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Park, Gee Young; Jung, H. S.; Ryu, J. S.; Park, C

2005-06-15

As the digital technology is being developed drastically, it is being applied to various industrial instrumentation and control (I and C) fields. In the nuclear power plants, I and C systems are also being installed by digital systems replacing their corresponding analog systems installed previously. There had been I and C systems constructed by analog technology especially for the reactor protection system in the research reactor HANARO. Parallel to the pace of the current trend for digital technology, it is desirable that all I and C systems including the safety critical and non-safety systems in an advanced research reactor is to be installed based on the computer based system. There are many attractable features in using digital systems against existing analog systems in that the digital system has a superior performance for a function and it is more flexible than the analog system. And any fruit gained from the newly developed digital technology can be easily incorporated into the existing digital system and hence, the performance improvement of a computer based system can be implemented conveniently and promptly. Moreover, the capability of high integrity in electronic circuits reduces the electronic components needed to construct the processing device and makes the electronic board simple, and this fact reveals that the hardware failure itself are unlikely to occur in the electronic device other than some electric problems. Balanced the fact mentioned above are the roles and related issues of the software loaded on the digital integrated hardware. Some defects in the course of software development might induce a severe damage on the computer system and plant systems and therefore it is obvious that comprehensive and deep considerations are to be placed on the development of the software in the design of I and C system for use in an advanced research reactor. The work investigates the domestic and international standards on the classifications of digital

Semiconductor research with reactor neutrons

International Nuclear Information System (INIS)

Kimura, Itsuro

1992-01-01

Reactor neutrons play an important role for characterization of semiconductor materials as same as other advanced materials. On the other hand reactor neutrons bring about not only malignant irradiation effects called radiation damage, but also useful effects such as neutron transmutation doping and defect formation for opto-electronics. Research works on semiconductor materials with the reactor neutrons of the Kyoto University Reactor (KUR) are briefly reviewed. In this review, a stress is laid on the present author's works. (author)

Development of steady thermal-hydraulic analysis code for China advanced research reactor

International Nuclear Information System (INIS)

Tian Wenxi; Qiu Suizheng; Guo Yun; Su Guanghui; Jia Dounan; Liu Tiancai; Zhang Jianwei

2006-01-01

A multi-channel model steady-state thermal-hydraulic analysis code was developed for China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed flow distribution in the core was obtained. The result shows that the structure size plays the most important role in flow distribution and the influence of core power could be neglected under single-phase flow. The temperature field of fuel element under unsymmetrical cooling condition was also obtained, which is necessary for the further study such as stress analysis etc. of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of hot channel was carried out and it is proved that all thermal-hydraulic parameters accord with the Safety Regulation of CARR. (authors)

Advanced reactor experimental facilities

International Nuclear Information System (INIS)

Amri, A.; Papin, J.; Uhle, J.; Vitanza, C.

2010-01-01

For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

Industrial structure at research reactor suppliers

International Nuclear Information System (INIS)

Roegler, H.-J.; Bogusch, E.; Friebe, T.

2001-01-01

Due to the recent joining of the forces of Framatome S. A. from France and the Nuclear Division of Siemens AG Power Generation (KWU) from Germany to a Joint Venture named Framatome Advanced Nuclear Power S.A.S., the issue of the necessary and of the optimal industrial structure for nuclear projects as a research reactor is, was discussed internally often and intensively. That discussion took place also in the other technical fields such as Services for NPPs but also in the field of interest here, i. e. Research Reactors. In summarizing the statements of this presentation one can about state that: Research Reactors are easier to build than NPPs, but not standardised; Research Reactors need a wide spectrum of skills and experiences; to design and build Research Reactors needs an experienced team especially in terms of management and interfaces; Research Reactors need background from built reference plants more than from operating plants; Research Reactors need knowledge of suitable experienced subsuppliers. Two more essential conclusions as industry involved in constructing and upgrading research reactors are: Research Reactors by far are more than a suitable core that generates a high neutron flux; every institution that designs and builds a Research Reactor lacks quality or causes safety problems, damages the reputation of the entire community

Advanced research reactor fuel development

Energy Technology Data Exchange (ETDEWEB)

Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

2000-05-01

The fabrication technology of the U{sub 3}Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U{sub 3}Si{sub 2} dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U{sub 3}Si{sub 2} fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 {approx} 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The {gamma}-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U{sub 3}Si{sub 2}. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano

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

Study for improvement of performance of the test and research reactors

Energy Technology Data Exchange (ETDEWEB)

Sakurai, Fumio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1999-03-01

Current utilization needs for the test and research reactors become more advanced and diversified along with the advance of nuclear science and technology. Besides, the requested safety for the research and test reactors grows strictly every year as well as a case of the power reactors. Under this circumstance, every effort to improve reactor performance including its safety is necessary to be sustained for allowing more effective utilization of the test and research reactors as experimental apparatus for advanced researches. In this study, the following three themes i.e., JMTR high-performance fuel element, evaluation method of fast neutron irradiation dose in the JMTR, evaluation method of performance of siphon break valve as core covering system for water-cooled test and research reactors, were investigated respectively from the views of improvement of core performance as a neutron source, utilization performance as an experimental apparatus, and safety as a reactor plant. (author)

Introduction of advanced pressurized water reactors in France

International Nuclear Information System (INIS)

Millot, J.P.; Nigon, M.; Vitton, M.

1988-01-01

Designed >30 yr ago, pressurized water reactors (PWRs) have evolved well to match the current safety, operating, and economic requirements. The first advanced PWR generation, the N4 reactor, is under construction with 1992 as a target date for commercial operation. The N4 may be considered to be a technological outcome of PWR evolution, providing advances in the fields of safety, man/machine interfaces, and load flexibility. As a step beyond N4, a second advanced PWR generation is presently under definition with, as a main objective, a greater ability to cope with the possible deterioration of the natural uranium market. In 1986, Electricite de France (EdF) launched investigations into the possible characteristics of this advanced PWR, called REP-2000 (PWR-2000: the reactor for the next century). Framatome joined EdF in 1987 but had been working on a new tight-lattice reactor. Main options are due by 1988; preliminary studies will begin and, by 1990, detailed design will proceed with the intent of firm commitments for the first unit by 1995. Commissioning is planned in the early years of the next century. This reactor type should be either an improved version of the N4 reactor or a spectral shift convertible reactor (RCVS). Through research and development efforts, Framatome, Commissariat a l'Energie Atomique (CEA), and EdF are investigating the physics of fuel rod tight lattices including neutronics, thermohydraulics, fuel behavior, and reactor mechanics

Development of inherent core technologies for advanced reactor

International Nuclear Information System (INIS)

Kim, Keung Koo; Noh, J.M.; Hwang, D.H.

1999-03-01

Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

Development of inherent core technologies for advanced reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Keung Koo; Noh, J.M.; Hwang, D.H. [and others

1999-03-01

Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

Advanced Reactor Technologies - Regulatory Technology Development Plan (RTDP)

Energy Technology Data Exchange (ETDEWEB)

Moe, Wayne L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-08-23

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory

Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

Energy Technology Data Exchange (ETDEWEB)

Moe, Wayne Leland [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-05-01

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory

The Advanced Neutron Source (ANS) project: A world-class research reactor facility

International Nuclear Information System (INIS)

Thompson, P.B.; Meek, W.E.

1993-01-01

This paper provides an overview of the Advanced Neutron Source (ANS), a new research facility being designed at Oak Ridge National Laboratory. The facility is based on a 330 MW, heavy-water cooled and reflected reactor as the neutron source, with a thermal neutron flux of about 7.5x10 19 m -2 ·sec -1 . Within the reflector region will be one hot source which will serve 2 hot neutron beam tubes, two cryogenic cold sources serving fourteen cold neutron beam tubes, two very cold beam tubes, and seven thermal neutron beam tubes. In addition there will be ten positions for materials irradiation experiments, five of them instrumented. The paper touches on the project status, safety concerns, cost estimates and scheduling, a description of the site, the reactor, and the arrangements of the facilities

Advanced In-Pile Instrumentation for Materials Testing Reactors

Science.gov (United States)

Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T. C.; Chase, B. M.; Davis, K. L.; Palmer, A. J.; Schley, R. S.

2014-08-01

The U.S. Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified, and the progress of other development efforts is summarized. As reported in this paper, INL researchers are currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating `advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers, are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors.

Digital control system of advanced reactor

International Nuclear Information System (INIS)

Peng Huaqing; Zhang Rui; Liu Lixin

2001-01-01

This article produced the Digital Control System For Advanced Reactor made by NPIC. This system uses Siemens SIMATIC PCS 7 process control system and includes five control system: reactor power control system, pressurizer level control system, pressurizer pressure control system, steam generator water level control system and dump control system. This system uses three automatic station to realize the function of five control system. Because the safety requisition of reactor is very strict, the system is redundant. The system configuration uses CFC and SCL. the human-machine interface is configured by Wincc. Finally the system passed the test of simulation by using RETRAN 02 to simulate the control object. The research solved the key technology of digital control system of reactor and will be very helpful for the nationalization of digital reactor control system

Office of Nuclear Regulatory Research summary of advanced reactors activities, June 4, 2001

International Nuclear Information System (INIS)

2001-01-01

Pre-application interactions with potential licensee applicants will help NRC prepare for future submittals, through the development of the infrastructure necessary for licensing application reviews. RES has the lead for non-LWR advanced reactor pre-application initiatives and longer-range new technology initiatives. An advanced reactor group has been formed in REAHFB, and is currently performing a pre-application review of Exelon's Pebble Bed Modular Reactor. Recent industry requests for future pre application interaction include General Atomics' Gas Turbine-Modular Helium Reactor (GT-MHR) and Westinghouse International Reactor Innovative and Secure (IRIS) design. RES advanced reactors activities also include participation as an observer in DOE's Generation IV initiative. Pre-Application review objectives include the development of regulatory guidance, licensing approach, and technology-basis expectations for licensing advanced designs, including identifying significant technology, design, safety, licensing and policy issues that would need to be addressed in the licensing process. The presentation described the pre-application process for the Exelon PBMR. NRC first identifies additional information following topical meetings with Exelon, and Exelon formally documents and submits required topical Information. The staff then develops a preliminary assessment and drafts a response which is followed by stakeholder input and comments at a public workshop. Preliminary assessments are discussed with ACRS and ACNW, and Commission papers are written which provide staff positions and recommendations on proposed policy decisions. Some of the significant areas for the PBMR include: Process Issues, Legal and Financial Issues; Regulatory Framework; Fuel Performance and Qualification; Traditional Engineering Design (e.g, Nuclear, Thermal-Fluid, Materials); Fuel Cycle Safety Areas; PRA, SSC Safety Classification; PBMR Prototype Testing

Advanced Test Reactor National Scientific User Facility

International Nuclear Information System (INIS)

Marshall, Frances M.; Benson, Jeff; Thelen, Mary Catherine

2011-01-01

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

Advanced Test Reactor National Scientific User Facility

Energy Technology Data Exchange (ETDEWEB)

Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

2011-08-01

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

A study on the development program of the advanced marine reactors

International Nuclear Information System (INIS)

Kobayashi, H.; Sako, K.; Iida, H.; Yamaji, A.

1992-01-01

JAERI has formulated two attractive concepts of advanced marine reactors. One is 100 MWt MRX (Marine Reactor X) for an icebreaker and the other is 150 kWe DRX (Deep-sea Reactor X) for a deep sea research submersible. They adopt new technologies such as an integral type PWR, in-vessel type control rod drive mechanisms, a water-filled containment vessel and a passive decay heat removal system, which would enable to satisfy the essential requirements for marine reactors for next generation, i.e.; compact, light, highly passive safe and easy to operate. From now on, following conceptual design, the engineering design phase is going to start in order to advance the research and development of MRX and DRX further and to obtain the data necessary for the detail design and construction of the actual reactors. JAERI is studying on the program to develop the engineering design research on MRX and DRX, which consists mainly of the particularization of design, the data acquisition by experiments (synthetic hydrothermal dynamics experiments, fundamental tests related to passive core cooling and demonstration tests on reliability and operability), the development of particular components and the development of advanced design tools. (author)

Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

International Nuclear Information System (INIS)

Moe, Wayne Leland

2015-01-01

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a ''critical path'' for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain ''minimum'' levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial ''first step'' in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by

Utilization of research reactors

International Nuclear Information System (INIS)

1962-01-01

About 200 research reactors are now in operation in different parts of the world, and at least 70 such facilities, which are in advanced stages of planning and construction, should be critical within the next two or three years. In the process of this development a multitude of problems are being encountered in formulating and carrying out programs for the proper utilization of these facilities, especially in countries which have just begun or are starting their atomic energy work. An opportunity for scientific personnel from different Member States to discuss research reactor problems was given at an international symposium on the Programing and Utilization of Research Reactors organized by the Agency almost immediately after the General Conference session. Two hundred scientists from 35 countries, as well as from the European Nuclear Energy Agency and EURATOM, attended the meeting which was held in Vienna from 16 to 21 October 1961

Development status of metallic, dispersion and non-oxide advanced and alternative fuels for power and research reactors

International Nuclear Information System (INIS)

2003-09-01

The current thermal power reactors use less than 1% of the energy contained in uranium. Long term perspectives aiming at a better economical extraction of the potential supplied by uranium motivated the development of new reactor types and, of course, new fuel concepts. Most of them dated from the sixties including liquid metal cooled fast (FR) and high temperature gas cooled (HTGR) reactors. Unfortunately, these impulses slowed down during the last twenty years; nuclear energy had to face political and consensus problems, in particular in the United States of America and in Europe, resulting from the consequences of the TMI and Chernobyl accidents. Good economical results obtained by the thermal power reactors also contributed to this process. During the last twenty years mainly France, India, Japan and the Russian Federation have maintained a relatively high level of technological development with appropriate financial items, in particular, in fuel research for the above mentioned reactor types. China and South Africa are now progressing in development of FR/HTGR and HTGR technologies, respectively. The purpose of this report is not only to summarise knowledge accumulated in the fuel research since the beginning of the sixties. This subject has been well covered in literature up to the end of the eighties. This report rather concentrates on the 'advanced fuels 'for the current different types of reactors including metallic, carbide and nitride fuels for fast reactors, so-called 'cold' fuels and fuels to burn excessive ex-weapons plutonium in thermal power reactors, alternative fuels for small size and research reactors. Emphasis has been put on the aspects of fabrication and irradiation behaviour of these fuels; available basic data concerning essential properties that help to understand the phenomena have been mentioned as well. This report brings complementary information to the earlier published monographs and concerns developments carried out after the early

Safety of research reactors (Design and Operation)

International Nuclear Information System (INIS)

Dirar, H. M.

2012-06-01

The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(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)

Assessment of Sensor Technologies for Advanced Reactors

Energy Technology Data Exchange (ETDEWEB)

Korsah, Kofi [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vlim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Kisner, Roger A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Britton, Jr, Charles L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wootan, D. W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Anheier, Jr, N. C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Diaz, A. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, E. H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chien, H. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Sheen, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States); Gopalsami, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Heifetz, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Tam, S. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Park, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Upadhyaya, B. R. [Univ. of Tennessee, Knoxville, TN (United States); Stanford, A. [Univ. of Tennessee, Knoxville, TN (United States)

2016-10-01

Sensors and measurement technologies provide information on processes, support operations and provide indications of component health. They are therefore crucial to plant operations and to commercialization of advanced reactors (AdvRx). This report, developed by a three-laboratory team consisting of Argonne National Laboratory (ANL), Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL), provides an assessment of sensor technologies and a determination of measurement needs for AdvRx. It provides the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program and contributes to the design and implementation of AdvRx concepts.

Advanced Test Reactor National Scientific User Facility Partnerships

International Nuclear Information System (INIS)

Marshall, Frances M.; Allen, Todd R.; Benson, Jeff B.; Cole, James I.; Thelen, Mary Catherine

2012-01-01

In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin

Advanced converters and reactors

International Nuclear Information System (INIS)

Haefele, W.; Kessler, G.

1984-01-01

As Western Europe and most countries of the Asia-Pacific region (except Australia) have only small natural uranium resources, they must import nuclear fuel from the major uranium supplier countries. The introduction of advanced converter and breeder reactor technology allows a fuel utilization of a factor of 4 to 100 higher than with present low converters (LWRs) and will make uranium-importing countries less vulnerable to price jumps and supply stops in the uranium market. In addition, breeder-reactor technology will open up a potential that can cover world energy requirements for several thousand years. The enormous development costs of advanced converter and breeder technologies can probably be raised only by highly industrialized countries. Those highly industrialized countries that have little or no uranium resources (Western Europe, Japan) will probably be the first to introduce this advanced reactor technology on a commercial scale. A number of small countries and islands will need only small power reactors with inherent safety capabilities, especially in the beginning of their nuclear energy programs. For economic reasons, the fuel cycle services should come from large reprocessing centers of countries having sufficiently large nuclear power programs or from international fuel cycle centers. (author)

Design of a multipurpose research reactor

International Nuclear Information System (INIS)

Sanchez Rios, A.A.

1990-01-01

The availability of a research reactor is essential in any endeavor to improve the execution of a nuclear programme, since it is a very versatile tool which can make a decisive contribution to a country's scientific and technological development. Because of their design, however, many existing research reactors are poorly adapted to certain uses. In some nuclear research centres, especially in the advanced countries, changes have been made in the original designs or new research prototypes have been designed for specific purposes. These modifications have proven very costly and therefore beyond the reach of developing countries. For this reason, what the research institutes in such countries need is a single sufficiently versatile nuclear plant capable of meeting the requirements of a nuclear research programme at a reasonable cost. This is precisely what a multipurpose reactor does. The Mexican National Nuclear Research Institute (ININ) plans to design and build a multipurpose research reactor capable at the same time of being used for the development of reactor design skills and for testing nuclear materials and fuels, for radioisotopes production, for nuclear power studies and basic scientific research, for specialized training, and so on. For this design work on the ININ Multipurpose Research Reactor, collaborative relations have been established with various international organizations possessing experience in nuclear reactor design: Atomehnergoeksport of the USSR: Atomic Energy of Canada Limited (AECL); General Atomics (GA) of the USA; and Japan Atomic Energy Research Institute

Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

Energy Technology Data Exchange (ETDEWEB)

Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

2007-06-30

This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

MIT research reactor. Power uprate and utilization

Energy Technology Data Exchange (ETDEWEB)

Hu, Lin-Wen [Nuclear Reactor Laboratory, Massachusetts Inst. of Technology, Cambridge, MA (United States)

2012-03-15

The MIT Research Reactor (MITR) is a university research reactor located on MIT campus. and has a long history in supporting research and education. Recent accomplishments include a 20% power rate to 6 MW and expanding advanced materials fuel testing program. Another important ongoing initiative is the conversion to high density low enrichment uranium (LEU) monolithic U-Mo fuel, which will consist of a new fuel element design and power increase to 7 MW. (author)

Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

Energy Technology Data Exchange (ETDEWEB)

Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1996-09-01

An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

Fuel, structural material and coolant for an advanced fast micro-reactor

International Nuclear Information System (INIS)

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

2011-01-01

The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials. (author)

NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapter 1, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

Advanced methods in teaching reactor physics

International Nuclear Information System (INIS)

Snoj, Luka; Kromar, Marjan; Zerovnik, Gasper; Ravnik, Matjaz

2011-01-01

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software.

Advanced methods in teaching reactor physics

Energy Technology Data Exchange (ETDEWEB)

Snoj, Luka, E-mail: luka.snoj@ijs.s [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Kromar, Marjan, E-mail: marjan.kromar@ijs.s [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Zerovnik, Gasper, E-mail: gasper.zerovnik@ijs.s [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Ravnik, Matjaz [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

2011-04-15

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software.

Advances by the Integral Fast Reactor Program

International Nuclear Information System (INIS)

Lineberry, M.J.; Pedersen, D.R.; Walters, L.C.; Cahalan, J.E.

1991-01-01

The advances by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, improved passive safety, and the development of a prototype fuel cycle facility. 14 refs

Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

Advanced gas-cooled reactors (AGR)

Energy Technology Data Exchange (ETDEWEB)

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

1981-01-15

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

Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor

Institute of Scientific and Technical Information of China (English)

TIAN Wenxi; QIU Suizheng; GUO Yun; SU Guanghui; JIA Dounan; LIU Tiancai; ZHANG Jianwei

2007-01-01

A multi-channel model steady-state thermalhydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under singlephase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. Of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the "Safety design regulation of CARR".

Advanced In-pile Instrumentation for Material and Test Reactors

International Nuclear Information System (INIS)

Rempe, J.L.; Knudson, D.L.; Daw, J.E.; Unruh, T.C.; Chase, B.M.; Davis, K.L.; Palmer, A.J.; Schley, R.S.

2013-06-01

The US Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified; and the progress of other development efforts is summarized. As reported in this paper, INL staff is currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating 'advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors. (authors)

Recent BWR fuel management reactor physics advances

International Nuclear Information System (INIS)

Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

1982-01-01

Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

Advanced reactor development for non-electric applications

International Nuclear Information System (INIS)

Chang, M.H.; Kim, S.H.

1996-01-01

Advance in the nuclear reactor technology achieved through nuclear power programs carried out in the world has led nuclear communities to direct its attention to a better and peaceful utilization of nuclear energy in addition to that for power generation. The efforts for non-electric application of nuclear energy has been pursued in a limited number of countries in the world for their special needs. However, those needs and the associated efforts contributed largely to the development and practical realization of advanced reactors characterized by highly improved reactor safety and reliability by deploying the most up-to-date safety technologies. Due mainly to the special purpose of utilization, economic reasons and ease in implementation of new advanced technologies, small and medium reactors have become a major stream in the reactor developments for non-electric applications. The purpose of this paper is to provide, to the interested nuclear society, the overview of the development status and design characteristics of selected advanced nuclear reactors previously developed and/or currently under development specially for non-electric applications. Major design technologies employed in those reactors to enhance the reactor safety and reliability are reviewed to present the underlying principles of the design. Along with the overview, this paper also introduces a development program and major design characteristics of an advanced integral reactor (SMART) for co-generation purpose currently under conceptual development in Korea. (author)

Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

Instrumentation to Enhance Advanced Test Reactor Irradiations

Energy Technology Data Exchange (ETDEWEB)

J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

2009-09-01

The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

Instrumentation to Enhance Advanced Test Reactor Irradiations

International Nuclear Information System (INIS)

Rempe, J.L.; Knudson, D.L.; Condie, K.G.; Daw, J.E.; Taylor, S.C.

2009-01-01

The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

Status of advanced light water reactor designs 2004

International Nuclear Information System (INIS)

2004-05-01

The report is intended to be a source of reference information for interested organizations and individuals. Among them are decision makers of countries considering implementation of nuclear power programmes. Further, the report is addressed to government officials with an appropriate technical background and to research institutes of countries with existing nuclear programmes that wish to be informed on the global status in order to plan their nuclear power programmes including both research and development efforts and means for meeting future. The future utilization of nuclear power worldwide depends primarily on the ability of the nuclear community to further improve the economic competitiveness of nuclear power plants while meeting stringent safety requirements. The IAEA's activities in nuclear power technology development include the preparation of status reports on advanced reactor designs to provide all interested IAEA Member States with balanced and objective information on advances in nuclear plant technology. In the field of light water reactors, the last status report published by the IAEA was 'Status of Advanced Light Water Cooled Reactor Designs: 1996' (IAEA-TECDOC-968). Since its publication, quite a lot has happened: some designs have been taken into commercial operation, others have achieved significant steps toward becoming commercial products, including certification from regulatory authorities, some are in a design optimization phase to reduce capital costs, development for other designs began after 1996, and a few designs are no longer pursued by their promoters. With this general progress in mind, on the advice and with the support of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for Light Water Reactors (LWRs), the IAEA has prepared this new status report on advanced LWR designs that updates IAEA-TECDOC-968, presenting the various advanced LWR designs in a balanced way according to a common outline

Preliminary design concepts of an advanced integral reactor

International Nuclear Information System (INIS)

Moon, Kap S.; Lee, Doo J.; Kim, Keung K.; Chang, Moon H.; Kim, Si H.

1997-01-01

An integral reactor on the basis of PWR technology is being conceptually developed at KAERI. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts of the reactor to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway for confirming the technical adoption of those concepts to the rector design. The power output of the reactor will be in the range of 100MWe to 600MWe which is relatively small compared to the existing loop type reactors. The detailed analysis to assure the design concepts is in progress. (author). 3 figs, 1 tab

French experience in design, operation and revamping of nuclear research reactors, in support of advanced reactors development

International Nuclear Information System (INIS)

Barre, B.; Bergeonneau, P.; Merchie, F.; Minguet, J.L.; Rousselle, P.

1996-01-01

The French nuclear program is strongly based on the R and D work performed in the CEA nuclear research centers and particularly on the various experimental programs carried out in its research reactors in the frame of cooperative actions between the Commissariat a l'Energie Atomique (CEA), Framatome and Electricite de France (EDF). Several types of research reactors have been built by Technicatome and CEA to carry out successfully this considerable R and D work on fuels and materials, among them the socalled Materials Testing Reactors (MTR) SILOE (35 MW) and OSIRIS (70 MW) which are indeed very well suited for technological irradiations. Their simple and flexible design and the large irradiation space available around the core, the SILOE and OSIRIS reactors can be shared by several types of applications such as fuel and material testings for nuclear power plants, radioisotopes production, silicon doping and fundamental research. It is worthwhile recalling that Technicatome and CEA have also built research reactors fully dedicated to safety experimental studies, such as the CABRI, SCARABEE and PHEBUS reactors at Cadarache, and others dedicated to fundamental research such as ORPHEE (14 MW) and the Reacteur a Haut Flux -High Flux Reactor- (RHF 57 MW). This paper will present some of the most significant conceptual and design features of all these reactors as well as the main improvements brought to most of them in the last years. Based on this wide experience, CEA and Technicatome have specially designed for export a new multipurpose research reactor named SIRIUS, with two versions depending on the utilization spectrum and the power range (5 MW to 30 MW). At last, CEA has recently launched the preliminary project study of a new MTR, the Jules Horowitz Reactor, to meet the future needs of fuels and materials irradiations in the next 4 or 5 decades, in support of the French long term nuclear power program. (J.P.N.)

Materials science research for sodium cooled fast reactors

Indian Academy of Sciences (India)

The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective ...

Thermohydraulic and safety analysis on China advanced research reactor under station blackout accident

International Nuclear Information System (INIS)

Tian Wenxi; Qiu Suizheng; Su Guanghui; Jia Dounan; Liu Xingmin; Zhang Jianwei

2007-01-01

A thermohydraulic and safety analysis code-TSACC has been developed using Fortran90 language to evaluate the transient thermohydraulic behavior of the China advanced research reactor (CARR) under station blackout accident (SBA). For the development of TSACC, a series of corresponding mathematical and physical models were applied. Point reactor neutron kinetics model was adopted for solving the reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional correlations were supplied. The usual finite difference method was abandoned and the integral technique was adopted to evaluate the temperature field of the plate type fuel elements. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behavior of the CARR. The computational result of TSACC showed the adequacy of the safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of RELAP5/MOD3 and a good agreement was obtained. The adoption of modular programming techniques enables TASCC to be applied to other reactors by easily modifying the corresponding function modules

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

International Nuclear Information System (INIS)

1976-01-01

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

The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

International Nuclear Information System (INIS)

Allen, T.R.; Benson, J.B.; Foster, J.A.; Marshall, F.M.; Meyer, M.K.; Thelen, M.C.

2009-01-01

To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

Advanced Reactor Development in the United States

Energy Technology Data Exchange (ETDEWEB)

Giessing, D. F.; Griffith, J. D.; McGoff, D. J.; Rosen, Sol [U. S. Department of Energy, Texas (United States)

1990-04-15

In the United States, three technologies are employed for the new generation of advanced reactors. These technologies are Advanced Light Water Reactors (A LWRs) for the 1990s and beyond, the Modular High Temperature Gas Reactor (M HTGR) for commercial use after the turn of the century, and Liquid Metal Reactors (LWRs) to provide energy production and to convert reactor fission waste to a more manageable waste product. Each technology contributes to the energy solution. Light Water Reactors For The 1990s And Beyond--The U. S. Program The economic and national security of the United States requires a diversified energy supply base built primarily upon adequate, domestic resources that are relatively free from international pressures. Nuclear energy is a vital component of this supply and is essential to meet current and future national energy demands. It is a safe, economically continues to contribute to national energy stability, and strength. The Light Water Reactor (LWR) has been a major and successful contributor to the electrical generating needs of many nations throughout the world. It is being counted upon in the United States as a key to revitalizing nuclear energy option in the 1990s. In recent years, DOE joined with the industry to ensure the availability and future viability of the LWR option. This national program has the participation of the Nation's utility industry, the Electric Power Research Institute (EPRI), and several of the major reactor manufacturers and architect-engineers. Separate but coordinated parts of this program are managed by EPRI and DOE.

An autonomous control framework for advanced reactors

Directory of Open Access Journals (Sweden)

Richard T. Wood

2017-08-01

Full Text Available Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors.

An autonomous control framework for advanced reactors

International Nuclear Information System (INIS)

Wood, Richard T.; Upadhyaya, Belle R.; Floyd, Dan C.

2017-01-01

Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors

An autonomous control framework for advanced reactors

Energy Technology Data Exchange (ETDEWEB)

Wood, Richard T.; Upadhyaya, Belle R.; Floyd, Dan C. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

2017-08-15

Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors.

The Budapest research reactor as an advanced research facility for the early 21st century

International Nuclear Information System (INIS)

Vidovszky, I.

2001-01-01

The Budapest Research Reactor, Hungary's first nuclear facility was originally put into operation in 1959. The reactor serves for: basic and applied research, technological and commercial applications, education and training. The main goal of the reactor is to serve neutron research. This unique research possibility is used by a broad user community of Europe. Eight instruments for neutron scattering, radiography and activation analyses are already used, others (e.g. time of flight spectrometer, neutron reflectometer) are being installed. The majority of these instruments will get a much improved utilization when the cold neutron source is put into operation. In 1999 the Budapest Research Reactor was operated for 3129 full power hours in 14 periods. The normal operation period took 234 hours (starting Monday noon and finishing Thursday morning). The entire production for the year 1999 was 1302 MW days. This is a slightly reduced value, due to the installation of the cold neutron source. For the year 2000 a somewhat longer operation is foreseen (near to 4000 hours), as the cold neutron source will be operational. The operation of the reactor is foreseen at least up to the end of the first decade of the 21 st century. (author)

New or improved computational methods and advanced reactor design

International Nuclear Information System (INIS)

Nakagawa, Masayuki; Takeda, Toshikazu; Ushio, Tadashi

1997-01-01

Nuclear computational method has been studied continuously up to date, as a fundamental technology supporting the nuclear development. At present, research on computational method according to new theory and the calculating method thought to be difficult to practise are also continued actively to find new development due to splendid improvement of features of computer. In Japan, many light water type reactors are now in operations, new computational methods are induced for nuclear design, and a lot of efforts are concentrated for intending to more improvement of economics and safety. In this paper, some new research results on the nuclear computational methods and their application to nuclear design of the reactor were described for introducing recent trend of the nuclear design of the reactor. 1) Advancement of the computational method, 2) Reactor core design and management of the light water reactor, and 3) Nuclear design of the fast reactor. (G.K.)

Comparative Study on Research Reactor Design Requirements between IAEA and Korea

International Nuclear Information System (INIS)

Chang, Won Joon; Yune, Young Gill; Song, Myung Ho; Cho, Seung Ho

2013-01-01

This study has identified the gaps in the safety requirements for design of research reactors of Korea comparing with those of the IAEA. The review results showed that the gaps have arisen mainly from the following aspects: - The differences in the characteristics of design and operation between power reactor and research reactor - Enhancement of the level of safety of nuclear reactor facility - Consideration of advanced safety technologies. The review results will be utilized to reflect the IAEA safety requirements for design of research reactors into those of Korea, which will contribute to enhancing the level of safety and improving the technical standards of research reactors of Korea. The IAEA safety standards encompass international consensus to strengthen the nuclear safety and to reflect the latest advancement of nuclear safety technologies. Also, they provide reliable means to ensure the effective fulfillment of obligations under the various international safety conventions. Many countries have adopted the IAEA safety standards as their national standards in nuclear regulations. Since Korea has exported research reactor technologies abroad these days and will continue to export them in the future, it is desirable to harmonize domestic safety requirements for research reactor with those of the IAEA. The KINS (Korea Institute of Nuclear Safety) has performed a review of the IAEA safety requirements for design of research reactors comparing with those of Korea. The purpose of this comparative study is to harmonize the safety requirements for the design of research reactors of Korea with those of the IAEA as a member state of the IAEA, and to encompass global efforts to enhance the nuclear safety and to reflect the latest advancement of nuclear safety technologies into the safety requirements for the design of research reactors of Korea. Design requirements for structures, systems, and components of research reactors important to safety, which are required to

Advances in fusion reactor design

International Nuclear Information System (INIS)

Baker, C.C.

1987-01-01

The author addresses the tokamak as a power reactor. Contrary to popular opinion, there are still a few people that think a tokamak might make a good fusion power reactor. In thinking about advances in fusion reactor design, in the U.S., at least, that generally means advances relevant to the Starfire design. He reviews some of the features of Starfire. Starfire is the last major study done of the tokamak as a reactor in this country. It is now over eight years old in the sense that eight years ago was really the time in which major decisions were made as to its features. Starfire was a tokamak with a major radius of seven meters, about twice the linear dimensions of a machine like TIBER

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)

U.S. Department of Energy instrumentation and controls technology research for advanced small modular reactors

International Nuclear Information System (INIS)

Wood, Richard Thomas

2013-01-01

Instrumentation, controls, and human-machine interfaces (ICHMI) are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The U.S. Department of Energy (DOE) has recognized that ICHMI research, development, and demonstration (RD and D) is needed to resolve the technical challenges that may compromise the effective and efficient utilization of modern ICHMI technology and consequently inhibit realization of the benefits offered by expanded utilization of nuclear power. Consequently, key DOE programs have substantial ICHMI RD and D elements to their respective research portfolio. This article describes current ICHMI research to support the development of advanced small modular reactors. (author)

Method of advancing research and development of fast breeder reactors

International Nuclear Information System (INIS)

1988-01-01

In the long term plan of atomic energy development and utilization, fast breeder reactors are to be developed as the main of the future nuclear power generation in Japan, and when their development is advanced, it has been decided to positively aim at building up the plutonium utilization system using FBRs superior to the uranium utilization system using LWRs. Also it has been decided that the development of FBRs requires to exert incessant efforts for a considerable long period under the proper cooperation system of government and people, and as for its concrete development, hereafter the deliberation is to be carried out in succession by the expert subcommittee on FBR development projects of the Atomic Energy Commission. The subcommittee was founded in May, 1986, to deliberate on the long term promotion measures for FBR development, the measures for promoting the research and development, the examination of the basic specification of a demonstration FBR, the measures for promoting international cooperation, and other important matters. As the results of investigation, the situation around the development of FBRs, the fundamentals at the time of promoting the research and development, the subjects of the research and development and so on are reported. (Kako, I.)

Simulation study on the cold neutron guides in China advanced research reactor

International Nuclear Information System (INIS)

Guo Liping; Yang Tonghua; Wang Hongli; Sun Kai; Zhao Zhixiang

2003-01-01

The designs of the two cold neutron guides, CNG1 and CNG2, to be built in China advanced research reactor (CARR) are studied with Monte-Carlo simulation technique. The neutron flux density at the exit of the both guides can reach above 1 x10 9 cm -2 ·s -1 under the assumed flux spectrum of the cold neutron source. The transmission efficiency is 50% and 42%, and the maximum divergence is about 2.2 degree and 1.9 degree, respectively for CNG1 and CNG2. Neutron distribution along horizontal direction is quite uniform for both guides, with maximum fluctuation of less than 3%. Gravity can affect neutron distribution along vertical direction considerably

Results of a comparison study of advanced reactors

International Nuclear Information System (INIS)

Bueno de Mesquita, K.G.; Gout, W.; Heil, J.A.; Tanke, R.H.J.; Geevers, F.

1991-06-01

The PINK programme is a 4-year programme of five parties involved in nuclear energy in the Netherlands: GKN (operator of the Dodewaard plant), KEMA (Research institute of the Netherlands Utilities), ECN (Netherlands Energy Research Foundation), NUCON (Engineering and Contracting Company) and IRI Interfaculty Reactor Institute of the Delft University of Technology), to coordinate their efforts to intensify the nuclear competence of the industry, the utilities and the research and engineering companies. This programme is sponsored by the Ministry of Economic Affairs. The PINK programme consists of five parts. This report pertains to part 1 of the programme: comparison study of advanced reactors concerning the four so-called second-stage designs SBWR, AP600, SIR and CANDU, which, compared to the first-stage reactor designs, features increased use of passive safety systems and simplification. The objective of the current study is to compare these advanced reactor designs in order to provide comprehensive information for the PINK steering committee that is useful in the selection process of a design for further study and development work. In ch. 2 the main features of the four reactors are highlighted. In ch. 3 the most important safety features and the behaviour of the four reactors under accident situations are compared. Passive safety systems are identified and forgivingness is described and compared. Results of the preliminary probabilistic safety analysis are presented. Ch. 4 deals with the proven technology of the four concepts, ch. 5 with the Netherlands requirements, ch. 6 with commercial aspects, and ch. 7 with the fuel cycle and radioactive waste produced. In ch. 8 the costs are compared and finally in ch. 9 conclusions are drawn and recommendations are made. (author). 13 figs

NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapters 2--13, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

The current status of utilization of research reactors in China

International Nuclear Information System (INIS)

Luzheng, Yuan

2004-01-01

Seminars on utilization of research reactors were held to enhance experience exchanging among institutes and universities in China. The status of CARR (China Advanced Research Reactor) project is briefly described. The progress in BNCT program in China is introduced. (author)

Thermohydraulic relationships for advanced water cooled reactors and the role of the IAEA

International Nuclear Information System (INIS)

Badulescu, A.; Groeneveld, D.C.

2000-01-01

Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1998. It was included into the IAEA's Programme following endorsement in 1995 by the International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote International Information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water-cooled reactors. (authors)

Comparisons among different development ways of advanced reactors in China

International Nuclear Information System (INIS)

Guo Xingqu; Lin Jianwen; Wang Ruoli

1992-03-01

For the development of nuclear energy in the 21st century, China will select a new type reactor to develop, which will have higher fuel efficiency, high safety and better economics. The selection is among the types of FBR (fast breeder reactor), HTGR (high temperature gas-cooled reactor) and FFHR (fusion-fission hybrid reactor). Since the evaluation of advanced reactors involves many uncertain factors and the difficulty of quantization, both the AHP (analytic hierarchy process) method and expert consultation are adopted. Four aspects are taken in the norm system of AHP, i.e. safety, maturity of technology, economy and appropriateness. By using questionnaire method to experts and studying related documents, five types of advanced reactor are selected, i.e. oxide fueled FBR, metal fueled FBR, uranium fueled HTGR, U-Th fueled HTGR and FFBR. Their evaluation parameters are a comprehensively assessed and sorted. About 130 experts and professors who have been working in the research institutes and government agencies of nuclear field are asked to give their comments on the development of advanced reactors. The response rate of questionnaires is 86%, and the data collected are processed by computers. From the evaluation result of AHP method and expert consultation of the fast breeder reactor, especially, the metal fueled FBR, should have the priority in nuclear energy development in the 21st century in China

Research reactor usage at the Idaho National Engineering Laboratory in support of university research and education

International Nuclear Information System (INIS)

Woodall, D.M.; Dolan, T.J.; Stephens, A.G.

1990-01-01

The Idaho National Engineering Laboratory is a US Department of Energy laboratory which has a substantial history of research and development in nuclear reactor technologies. There are a number of available nuclear reactor facilities which have been incorporated into the research and training needs of university nuclear engineering programs. This paper addresses the utilization of the Advanced Reactivity Measurement Facility (ARMF) and the Coupled Fast Reactivity Measurement Facility (CFRMF) for thesis and dissertation research in the PhD program in Nuclear Science and Engineering by the University of Idaho and Idaho State University. Other reactors at the INEL are also being used by various members of the academic community for thesis and dissertation research, as well as for research to advance the state of knowledge in innovative nuclear technologies, with the EBR-II facility playing an essential role in liquid metal breeder reactor research. 3 refs

Advances in heavy water reactors

International Nuclear Information System (INIS)

1994-03-01

The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The Technical Committee Meeting (TCM) on Advances in Heavy Water Reactors was organized by the IAEA in the framework of the activities of the International Working Group on Advanced Technologies for Water Cooled Reactors (IWGATWR) and hosted by the Atomic Energy of Canada Limited. Sixty-five participants from nine countries (Canada, Czech Republic, India, German, Japan, Republic of Korea, Pakistan, Romania and USA) and the IAEA attended the TCM. Thirty-four papers were presented and discussed in five sessions. A separate abstract was prepared for each of these papers. All recommendations which were addressed by the participants of the Technical Committee meeting to the IWGATWR have been submitted to the 5th IWGATWR meeting in September 1993. They were reviewed and used as input for the preparation of the IAEA programme in the area of advanced water cooled reactors. This TCM was mainly oriented towards advances in HWRs and on projects which are now in the design process and under discussion. Refs, figs and tabs

Advanced burner test reactor preconceptual design report.

Energy Technology Data Exchange (ETDEWEB)

Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

2008-12-16

advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.

Rationalization and future planning for AECL's research reactor capability

International Nuclear Information System (INIS)

Slater, J.B.

1990-01-01

AECL's research reactor capability has played a crucial role in the development of Canada's nuclear program. All essential concepts for the CANDU reactors were developed and tested in the NRX and NRU reactors, and in parallel, important contributions to basic physics were made. The technical feasibility of advanced fuel cycles and of the organic-cooled option for CANDU reactors were also demonstrated in the two reactors and the WR-1 reactor. In addition, an important and growing radio-isotope production industry was established and marketed on a world-wide basis. In 1984, however, it was recognized that a review and rationalization of the research reactor capability was required. The commercial success of the CANDU reactor system had reduced the scope and size of the required development program. Limited research and development funding and competition from other research facilities and programs, required that the scope be reduced to a support basis essential to maintain strategic capability. Currently, AECL, is part-way through this rationalization program and completion should be attained during 1992/93 when the MAPLE reactor is operational and decisions on NRX decommissioning will be made. A companion paper describes some of the unique operational and maintenance problems which have resulted from this program and the solutions which have been developed. Future planning must recognize the age of the NRU reactor (currently 32 years) and the need to plan for eventual replacement. Strategy is being developed and supporting studies include a full technical assessment of the NRU reactor and the required age-related upgrading program, evaluation of the performance characteristics and costs of potential future replacement reactors, particularly the advanced MAPLE concept, and opportunities for international co-operation in developing mutually supportive research programs

Research for enhancing reactor safety

International Nuclear Information System (INIS)

1989-05-01

Recent research for enhanced reactor safety covers extensive and numerous experiments and computed modelling activities designed to verify and to improve existing design requirements. The lectures presented at the meeting report GRS research results and the current status of reactor safety research in France. The GRS experts present results concerning expert systems and their perspectives in safety engineering, large-scale experiments and their significance in the development and verification of computer codes for thermohydraulic modelling of safety-related incidents, the advanced system code ATHLET for analysis of thermohydraulic processes of incidents, the analysis simulator which is a tool for fast evaluation of accident management measures, and investigations into event sequences and the required preventive emergency measures within the German Risk Study. (DG) [de

Sodium fast reactor safety and licensing research plan - Volume II

International Nuclear Information System (INIS)

Ludewig, H.; Powers, D.A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.

2012-01-01

Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

Sodium fast reactor safety and licensing research plan. Volume II.

Energy Technology Data Exchange (ETDEWEB)

Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

2012-05-01

Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

Licensing of advanced reactors: Status report and perspective

International Nuclear Information System (INIS)

King, T.

1988-01-01

In July, 1986, the U.S. Nuclear Regulatory Commission issued a Policy State on the Regulation of Advanced Nuclear Power Plants. As part of this policy, advanced reactor designers were encouraged to interact with NRC [Nuclear Regulatory Commission] early in the design process to obtain feedback regarding licensing requirements for advanced reactors. Accordingly, the staff has been interacting with the Department of Energy (DOE) and its contractors on the review of three advanced reactor conceptual designs: one modular high temperature gas-cooled reactor (MHTGR) and two liquid metal reactors (LMRs). This paper provides a status of the NRC review effort, describes the key policy and technical issues resulting from our review and provides the current status and approach to the development of licensing guidance on each

Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

International Nuclear Information System (INIS)

Song, C. H.; Chung, M. K.; Park, C. K. and others

2005-04-01

The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved

Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Song, C. H.; Chung, M. K.; Park, C. K. and others

2005-04-15

The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.

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

Advanced light-water reactors

International Nuclear Information System (INIS)

Golay, M.W.; Todreas, N.E.

1990-01-01

Environmental concerns, economics and the earth's finite store of fossil fuels argue for a resuscitation of nuclear power. The authors think improved light-water reactors incorporating passive safety features can be both safe and profitable, but only if attention is paid to economics, effective management and rigorous training methods. The experience of nearly four decades has winnowed out designs for four basic types of reactor: the heavy-water reactor (HWR), the gas-cooled rector (GCR), the liquid-metal-cooled reactor (LMR) and the light-water reactor (LWR). Each design is briefly described before the paper discusses the passive safety features of the AP-600 rector, so-called because it employs an advanced pressurized water design and generates 600 MW of power

Nuclear research reactors

International Nuclear Information System (INIS)

1985-01-01

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

Development of advanced strain diagnostic techniques for reactor environments.

Energy Technology Data Exchange (ETDEWEB)

Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.; Hall, Aaron Christopher; Urrea, David Anthony,; Parma, Edward J.,

2013-02-01

The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding. During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.

Materials research with neutron beams from a research reactor

Energy Technology Data Exchange (ETDEWEB)

Root, J.; Banks, D. [Canadian Neutron Beam Centre, Chalk River Laboratories, Chalk River, Ontario (Canada)

2015-03-15

Because of the unique ways that neutrons interact with matter, neutron beams from a research reactor can reveal knowledge about materials that cannot be obtained as easily with other scientific methods. Neutron beams are suitable for imaging methods (radiography or tomography), for scattering methods (diffraction, spectroscopy, and reflectometry) and for other possibilities. Neutron-beam methods are applied by students and researchers from academia, industry and government to support their materials research programs in several disciplines: physics, chemistry, materials science and life science. The arising knowledge about materials has been applied to advance technologies that appear in everyday life: transportation, communication, energy, environment and health. This paper illustrates the broad spectrum of materials research with neutron beams, by presenting examples from the Canadian Neutron Beam Centre at the NRU research reactor in Chalk River. (author)

U.S. Research Program to Support Advanced Reactors and Fuel Cycle Options

International Nuclear Information System (INIS)

Lyons, Peter

2013-01-01

• In recognition of possible future needs, the U.S. will perform R&D on advanced reactor and fuel cycle technologies that could dramatically improve nuclear energy safety and performance; • Multifaceted approach to support R&D: - National labs; - Universities; - Industry; - International partners

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.

Analytical chemistry requirements for advanced reactors

International Nuclear Information System (INIS)

Jayashree, S.; Velmurugan, S.

2015-01-01

The nuclear power industry has been developing and improving reactor technology for more than five decades. Newer advanced reactors now being built have simpler designs which reduce capital cost. The greatest departure from most designs now in operation is that many incorporate passive or inherent safety features which require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. India is developing the Advanced Heavy Water Reactor (AHWR) in its plan to utilise thorium in nuclear power program

The advanced test reactor national scientific user facility advancing nuclear technology

International Nuclear Information System (INIS)

Allen, T.R.; Thelen, M.C.; Meyer, M.K.; Marshall, F.M.; Foster, J.; Benson, J.B.

2009-01-01

To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

Trends in advanced reactor development and the role of the IAEA

International Nuclear Information System (INIS)

Semenov, B.; Dastidar, P.; Kupitz, J.; Cleveland, J.; Goodjohn, A.

1992-01-01

This report discusses advanced reactors are being developed for all principal reactor types, i.e. the light and heavy water-cooled reactors, the liquid-metal-cooled reactors and the gas-cooled reactors. Some of these developments are primarily of an evolutionary nature, i.e. they represent improvements in component and system technology, and in construction and operating practices as a result of experience gained with presently operating plants. Other developments are also evolutionary but with some incorporation of innovative features such as providing passive systems for assuring continuous cooling for removal of decay heat from the reactor core. If there is a revival of nuclear power, which may be dictated by ecological and economical factors, advanced reactors now being developed could help to meet the large demand for new plants in developed and developing countries, not only for electricity generation, but also for district heating, desalination and for process heat. The IAEA, as the only global international governmental organization dealing with nuclear power, has promoted international information exchange and international cooperation between all countries with their own advanced nuclear power programmes and has offered assistance to countries with an interest in exploratory or research programmes. In the future the IAEA could play an even more-important role

Status of advanced nuclear reactor development in Korea

International Nuclear Information System (INIS)

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

2014-01-01

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

Safeguarding research reactors

International Nuclear Information System (INIS)

Powers, J.A.

1983-03-01

The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

Experimental research subject and renovation of chemical processing facility (CPF) for advanced fast reactor fuel reprocessing technology development

International Nuclear Information System (INIS)

Koyama, Tomozo; Shinozaki, Tadahiro; Nomura, Kazunori; Koma, Yoshikazu; Miyachi, Shigehiko; Ichige, Yoshiaki; Kobayashi, Tsuguyuki; Nemoto, Shin-ichi

2002-01-01

In order to enhance economical efficiency, environmental impact and nuclear nonproliferation resistance, the Advanced Reprocessing Technology, such as simplification and optimization of process, and applicability evaluation of the innovative technology that was not adopted up to now, has been developed for the reprocessing of the irradiated fuel taken out from a fast reactor. Renovation of the hot cell interior equipments, establishment and updating of glove boxes, installation of various analytical equipments, etc. in the Chemical Processing Facility (CPF) was done to utilize the CPF more positivity which is the center of the experimental field, where actual fuel can be used, for research and development towards establishment of the Advanced Reprocessing Technology development. The hot trials using the irradiated fuel pins of the experimental fast reactor 'JOYO' for studies on improved aqueous reprocessing technology, MA separation technology, dry process technology, etc. are scheduled to be carried out with these new equipments. (author)

Accident analysis in research reactors

International Nuclear Information System (INIS)

Adorni, M.; Bousbia-salah, A.; D'Auria, F.; Hamidouche, T.

2007-01-01

With the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Consequently, advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. The challenge today is to revisit the safety features of the existing nuclear plants and particularly research reactors in order to verify that the safety requirements are still met and - when necessary - to introduce some amendments not only to meet the new requirements but also to introduce new equipment from recent development of new technologies. The purpose of the present paper is to provide an overview of the accident analysis technology applied to the research reactor, with emphasis given to the capabilities of computational tools. (author)

Fast reactor research in Switzerland

International Nuclear Information System (INIS)

Brogli, R.; Hudina, M.; Pelloni, S.; Sigg, B.; Stanculescu, A.

1998-01-01

The small Swiss research program on fast reactors serves to further understanding of the role of LMFR for energy production and to convert radioactive waste to more environmentally benign forms. These activities are on the one hand the contribution to the comparison of advanced nuclear systems and bring on the other to our physical and engineers understanding. (author)

Advanced nuclear reactor types and technologies

International Nuclear Information System (INIS)

Ignatiev, V.; Devell, L.

1995-01-01

The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary

Advanced nuclear reactor types and technologies

Energy Technology Data Exchange (ETDEWEB)

Ignatiev, V [ed.; Feinberg, O; Morozov, A [Russian Research Centre ` Kurchatov Institute` , Moscow (Russian Federation); Devell, L [Studsvik Eco and Safety AB, Nykoeping (Sweden)

1995-07-01

The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

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

International Nuclear Information System (INIS)

1989-07-01

In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the 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 trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

A classification plan of design class for systems of an advanced research reactor

International Nuclear Information System (INIS)

Yoon, Doo Byung; Ryu, Jeong Soo

2005-01-01

Advanced Research Reactor(ARR) is being designed by KAERI since 2002. The final goal of the project is to develop a new and unique research reactor model which is superior in safety and economical aspects. The conceptual design for systems, structures, and components of the ARR will be completed by 2005. The basic design for the systems, structures, and components of the ARR will be performed from 2006. Based on the technical experiences on the design and operation of the HANARO, the ARR will be designed. It is necessary to classify the safety class, quality class, and seismic category for the systems, structures, and components. The objective of this work is to propose a classification plan of design class for systems, structures, and components of the ARR. To achieve this purpose, the revision status of the regulations that used as criteria for determining the design class of the systems, structures, and components of the HANARO were investigated. In addition, the present revision status of the codes and the standards that utilized for the design of the HANARO were investigated. Based on these investigations, the codes and the standards for the design of the systems, structures, and components of the ARR were proposed. The feasibility of the proposed classification plan will be verified by performing the conceptual and basic design of the systems, structures, and components of the ARR

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

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

Special issue on the "Consortium for Advanced Simulation of Light Water Reactors Research and Development Progress"

Science.gov (United States)

Turinsky, Paul J.; Martin, William R.

2017-04-01

In this special issue of the Journal of Computational Physics, the research and development completed at the time of manuscript submission by the Consortium for Advanced Simulation of Light Water Reactors (CASL) is presented. CASL is the first of several Energy Innovation Hubs that have been created by the Department of Energy. The Hubs are modeled after the strong scientific management characteristics of the Manhattan Project and AT&T Bell Laboratories, and function as integrated research centers that combine basic and applied research with engineering to accelerate scientific discovery that addresses critical energy issues. Lifetime of a Hub is expected to be five or ten years depending upon performance, with CASL being granted a ten year lifetime.

Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor

International Nuclear Information System (INIS)

Ott, Larry J.; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J.; Bevard, Bruce Balkcom

2009-01-01

The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

International Nuclear Information System (INIS)

2012-03-01

The role of natural circulation in advanced water cooled reactor design has been extended with the adoption of passive safety systems. Some designs utilize natural circulation to remove core heat during normal operation. Most passive safety systems used in evolutionary and innovative water cooled reactor designs are driven by natural circulation. The use of passive systems based on natural circulation can eliminate the costs associated with the installation, maintenance and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. Several IAEA Member States with advanced reactor development programmes are actively conducting investigations of natural circulation to support the development of advanced water cooled reactor designs with passive safety systems. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, in 2004 the IAEA initiated a coordinated research project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation. Three reports were published within the framework of this CRP. The first report (IAEA-TECDOC-1474) contains the material developed for the first IAEA training course on natural circulation in water cooled nuclear power plants. The second report (IAEA-TECDOC-1624) describes passive safety systems in a wide range of advanced water cooled nuclear power plant designs, with the goal of gaining insights into system design, operation and reliability. This third, and last, report summarizes the research studies completed by participating institutes during the CRP period.

Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

International Nuclear Information System (INIS)

Song, C. H.; Baek, W. P.; Chung, M. K.

2007-06-01

The objectives of the project are to study thermal hydraulic characteristics of advanced nuclear reactor system for evaluating key thermal-hydraulic phenomena relevant to new safety concepts. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. The Followings are main research topics: - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation-induced Thermal Mixing in a Pool - Development of Thermal-Hydraulic Models for Two-Phase Flow - Construction of T-H Data Base

Novelties in design and construction of the advanced reactors

International Nuclear Information System (INIS)

Acosta Ezcurra, T.; Garcia Rodriguez, B.M.

1996-01-01

The advanced pressurized water reactors (APWR), advanced boiling water reactors (ABWR), advanced liquid metal reactors (ALMR), and modular high temperature gas-cooled reactors (MHTGR), as well as heavy water reactors (AHWR), are analyzed taking into account those characteristics which make them less complex, but safer than their current homologous ones. This fact simplifies their construction which reduces completion periods and costs, increasing safety and protection of the plants. It is demonstrated how the accumulated operational experience allows to find more standardized designs with some enhancement in the material and component technology and thus achieve also a better use of computerized systems

Plant maintenance and advanced reactors issue, 2008

Energy Technology Data Exchange (ETDEWEB)

Agnihotri, Newal [ed.

2009-09-15

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Technologies of national importance, by Tsutomu Ohkubo, Japan Atomic Energy Agency, Japan; Modeling and simulation advances brighten future nuclear power, by Hussein Khalil, Argonne National Laboratory, Energy and desalination projects, by Ratan Kumar Sinha, Bhabha Atomic Research Centre, India; A plant with simplified design, by John Higgins, GE Hitachi Nuclear Energy; A forward thinking design, by Ray Ganthner, AREVA; A passively safe design, by Ed Cummins, Westinghouse Electric Company; A market-ready design, by Ken Petrunik, Atomic Energy of Canada Limited, Canada; Generation IV Advanced Nuclear Energy Systems, by Jacques Bouchard, French Commissariat a l'Energie Atomique, France, and Ralph Bennett, Idaho National Laboratory; Innovative reactor designs, a report by IAEA, Vienna, Austria; Guidance for new vendors, by John Nakoski, U.S. Nuclear Regulatory Commission; Road map for future energy, by John Cleveland, International Atomic Energy Agency, Vienna, Austria; and, Vermont's largest source of electricity, by Tyler Lamberts, Entergy Nuclear Operations, Inc. The Industry Innovation article is titled Intelligent monitoring technology, by Chris Demars, Exelon Nuclear.

The Conceptual Design for Tubular Fuel Assemblies of an Advanced Research Reactor

International Nuclear Information System (INIS)

Ryu, Jeong Soo; Dan, Ho Jin; Cho, Yeong Garp; Yoon, Doo Byung; Park, Cheol

2005-05-01

An Advanced Research Reactor(ARR) is being designed by KAERI since 2002. The final goal of the project is to develop a new and unique research reactor model which is superior in safety and economical aspects. In this work, the conceptual design for tubular fuel assemblies was carried out to enhance the previous model. The shape optimization of the cross section of the top guide was performed, and the swaging procedure in connecting fuel plates and stiffeners was developed. Moreover to reflect changes in number and size of fuel plates, related parts of the standard and the reduced fuel assemblies were redesigned. The top guide should suppress the vibration of the fuel assembly due to coolant and resist against material failures owing to fatigue and yield. In order to gain these design requirements, we have optimized the section profile of the top guide. To confirm manufacturing aspects, the swaging procedure was developed and its performance was tested. The results of tangential tensile test and axial compression test guaranteed that the fixing state between fuel plates and stiffeners is firm enough to hold each other. In addition, due to changes in number and size of fuel plates, the outer cross section of the fuel assembly was expanded and the diameter of the spacer tube was reduced. Reflecting these design changes, top/bottom guide, top guide cover, spring, spring cover, and receptacle were readjusted. Based on the technical experiences on the design and operation of the HANARO, the standard and the reduced fuel assemblies will be verified by performing various tests and analysis

Trends in advanced reactor development and the role of the IAEA

International Nuclear Information System (INIS)

Kupitz, J.

1992-01-01

Worldwide a tremendous amount of experience has been accumulated during development, licensing, construction and operation of nuclear power reactors. The experience forms a sound basis for further improvements. Nuclear programmes in many countries are addressing the development of advanced reactors which are intended to have better economics, higher reliability and improved safety in order to overcome the current concerns of nuclear power. Advanced reactors now being developed could help to meet the demand for nev plants in developed and developing countries, not only for electricity generation, but also for district heating, desalination and for process heat. This report discussed the role of IAEA, as the only global international governmental organization dealing with nuclear power, which promotes international information exchange and international cooperation between all countries with their own advanced nuclear power programmes and offers assistance to countries with an interest in exploratory or research programmes

Mirror Advanced Reactor Study (MARS)

International Nuclear Information System (INIS)

Logan, B.G.

1983-01-01

Progress in a two year study of a 1200 MWe commercial tandem mirror reactor (MARS - Mirror Advanced Reactor Study) has reached the point where major reactor system technologies are identified. New design features of the magnets, blankets, plug heating systems and direct converter are described. With the innovation of radial drift pumping to maintain low plug density, reactor recirculating power fraction is reduced to 20%. Dominance of radial ion and impurity losses into the halo permits gridless, circular direct converters to be dramatically reduced in size. Comparisons of MARS with the Starfire tokamak design are made

Enhanced in-pile instrumentation at the advanced test reactor

Energy Technology Data Exchange (ETDEWEB)

Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

2011-07-01

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

Enhanced In-Pile Instrumentation at the Advanced Test Reactor

Science.gov (United States)

Rempe, Joy L.; Knudson, Darrell L.; Daw, Joshua E.; Unruh, Troy; Chase, Benjamin M.; Palmer, Joe; Condie, Keith G.; Davis, Kurt L.

2012-08-01

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

Fuel rod bundles proposed for advanced pressure tube nuclear reactors

International Nuclear Information System (INIS)

Prodea, Iosif; Catana, Alexandru

2010-01-01

The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACR TM -1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, New York (United States)

2007-07-01

Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat.

Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan; Kim, Yeon Soo

2007-01-01

Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

Development of Digital MMIS for Research Reactors: Graded Approaches

Energy Technology Data Exchange (ETDEWEB)

Khalil ur, Rahman; Shin, Jin Soo; Heo, Gyun Young [Kyunghee University, Yongin (Korea, Republic of); Son, Han Seong [Joongbu University, Geumsan (Korea, Republic of); Kim, Young Ki; Park, Jae Kwan; Seo, Sang Mun; Kim, Yong Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-05-15

Though research reactors are small in size yet they are important in terms of industrial applications and R and D, educational purposes. Keeping the eye on its importance, Korean government has intention to upgrade and extend this industry. Presently, Korea is operating only HANARO at Korea Atomic Energy Research Institute (KAERI) and AGN-201K at Kyung Hee University (KHU), which are not sufficient to meet the current requirements of research and education. In addition, we need self-sufficiency in design and selfreliance in design and operation, as we are installing research reactors in domestic as well as foreign territories for instance Jordan. Based on these demands, KAERI and universities initiated a 5 year research project since December 2011 collaboratly, for the deep study of reactor core, thermal hydraulics, materials and instrumentation and control (I and C). This particular study is being carried out to develop highly reliable advanced digital I and C systems using a grading approach. It is worth mentioning that next generation research reactor should be equipped with advance state of the art digital I and C for safe and reliable operation and impermeable cyber security system that is needed to be devised. Moreover, human error is one of important area which should be linked with I and C in terms of Man Machine Interface System (MMIS) and development of I and C should cover human factor engineering. Presently, the digital I and C and MMIS are well developed for commercial power stations whereas such level of development does not exist for research reactors in Korea. Since the functional and safety requirements of research reactors are not so strict as commercial power plants, the design of digital I and C systems for research reactors seems to be graded based on the stringency of regulatory requirements. This paper was motivated for the introduction of those missions, so it is going to describe the general overview of digital I and C systems, the graded

Development of Digital MMIS for Research Reactors: Graded Approaches

International Nuclear Information System (INIS)

Khalil ur, Rahman; Shin, Jin Soo; Heo, Gyun Young; Son, Han Seong; Kim, Young Ki; Park, Jae Kwan; Seo, Sang Mun; Kim, Yong Jun

2012-01-01

Though research reactors are small in size yet they are important in terms of industrial applications and R and D, educational purposes. Keeping the eye on its importance, Korean government has intention to upgrade and extend this industry. Presently, Korea is operating only HANARO at Korea Atomic Energy Research Institute (KAERI) and AGN-201K at Kyung Hee University (KHU), which are not sufficient to meet the current requirements of research and education. In addition, we need self-sufficiency in design and selfreliance in design and operation, as we are installing research reactors in domestic as well as foreign territories for instance Jordan. Based on these demands, KAERI and universities initiated a 5 year research project since December 2011 collaboratly, for the deep study of reactor core, thermal hydraulics, materials and instrumentation and control (I and C). This particular study is being carried out to develop highly reliable advanced digital I and C systems using a grading approach. It is worth mentioning that next generation research reactor should be equipped with advance state of the art digital I and C for safe and reliable operation and impermeable cyber security system that is needed to be devised. Moreover, human error is one of important area which should be linked with I and C in terms of Man Machine Interface System (MMIS) and development of I and C should cover human factor engineering. Presently, the digital I and C and MMIS are well developed for commercial power stations whereas such level of development does not exist for research reactors in Korea. Since the functional and safety requirements of research reactors are not so strict as commercial power plants, the design of digital I and C systems for research reactors seems to be graded based on the stringency of regulatory requirements. This paper was motivated for the introduction of those missions, so it is going to describe the general overview of digital I and C systems, the graded

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

International Nuclear Information System (INIS)

1976-01-01

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

Shielding considerations for advanced space nuclear reactor systems

International Nuclear Information System (INIS)

Angelo, J.P. Jr.; Buden, D.

1982-01-01

To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO 2 ) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications

Trends in the design of advanced nuclear reactors

International Nuclear Information System (INIS)

Poong-Eil Juhn; Kupitz, Juergen

1996-01-01

Nuclear energy is an essentially unlimited energy source with the potential to provide energy in the form of electricity, district heat and process heat environmentally acceptable conditions. However, this potential will be realized only if nuclear power plants can meet the challenges of national safety requirements, economic competitiveness and public acceptance. Worldwide, a tremendous amount of experience has been accumulated during the development, licensing, construction and operation of nuclear power plants. This experience forms a sound basis for further improvements. Nuclear programmes in the IAEA Member States are addressing the development of advanced reactors, which are intended to have better economics, higher reliability and improved safety. The IAEA, as a global international governmental organization dealing with nuclear power, promotes international information exchange and international co-operation between all countries with their own advanced power programmes and offers assistance to countries with an interest in exploratory or research programmes. The paper gives an overview of global trends in the design of advanced nuclear reactors for electricity generation and heat production along with the role of IAEA. (author)

Development of a system model for advanced small modular reactors.

Energy Technology Data Exchange (ETDEWEB)

Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

2014-01-01

This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

Status of advanced technologies for CANDU reactors

International Nuclear Information System (INIS)

Lipsett, J.J.

1989-01-01

The future development of the CANDU reactor is a continuation of a successful series of reactors, the most recent of which are nine CANDU 6 Mk 1* units and four Darlington units. There are three projects underway that continue the development of the CANDU reactor. These new design projects flow from the original reactor designs and are a natural progression of the CANDU 6 Mk 1, two units of which are operating successfully in Canada, one each in Argentina and Korea, with five more being built in Rumania. These new design projects are known as: CANDU 6 Mk 2, an improved version of CANDU 6 Mk 1; CANDU 3, a small, advanced version of the CANDU 6 Mk 1; CANDU 6 Mk 3, a series of advanced CANDU reactors. A short description of modified versions of CANDU reactors is given in this paper. 5 figs

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

GE's advanced nuclear reactor designs

International Nuclear Information System (INIS)

Berglund, R.C.

1993-01-01

The excess of US electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14-point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other open-quotes enabling conditions.close quotes GE is participating in this national effort and GE's family of advanced nuclear power plants feature two reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the US and worldwide. Both possess the features necessary to do so safety, reliably, and economically

Jordan Research and Training Reactor (JRTR) Utilization Facilities

International Nuclear Information System (INIS)

Xoubi, N.

2013-01-01

Jordan Research and Training Reactor (JRTR) is a 5 MW light water open pool multipurpose reactor that serves as the focal point for Jordan National Nuclear Centre, and is designed to be utilized in three main areas: Education and training, nuclear research, and radioisotopes production and other commercial and industrial services. The reactor core is composed of 18 fuel assemblies, MTR plate type 19.75% enriched uranium silicide (U 3 Si 2 ) in aluminium matrix, and is reflected on all sides by beryllium and graphite. The reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45×10 14 cm -2 s -1 , and is controlled by a Hafnium control absorber rod and B 4 C shutdown rod. The reactor is designed to include laboratories and classrooms that will support the establishment of a nuclear reactor school for educating and training students in disciplines like nuclear engineering, reactor physics, radiochemistry, nuclear technology, radiation protection, and other related scientific fields where classroom instruction and laboratory experiments will be related in a very practical and realistic manner to the actual operation of the reactor. JRTR is designed to support advanced nuclear research as well as commercial and industrial services, which can be preformed utilizing any of its 35 experimental facilities. (author)

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

International Nuclear Information System (INIS)

Akino, Fujiyoshi

1988-01-01

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

Halden Reactor Project Workshop: Understanding Advanced Instrumentation and Controls Issues

International Nuclear Information System (INIS)

Beltracchi, L.

1991-01-01

A Halden Reactor Project Workshop on 'Understanding Advanced Instrumentation and Controls Issues' was held in Halden, Norway, during June 17-18, 1991. The objectives of the workshop were to (1) identify and prioritize the types of technical information that the Halden Project can produce to facilitate the development of man-machine interface guidelines and (2) to identify methods to effectively integrate and disseminate this information to signatory organizations. As a member of the Halden Reactor Project, the Nuclear Regulatory Commission (NRC) requested the workshop. This request resulted from the NRC's need for human factors guidelines for the evaluation of advanced instrumentation and controls. The Halden Reactor Project is a cooperative agreement among several countries belonging to the Organization for Economic Cooperation and Development (OECD). The US began its association with the Halden Project in 1958 through the Atomic Energy Commission. The project's activities are centered at the Halden heavy-water reactor and its associated man-machine laboratory in Halden, Norway. The research program conducted at Halden consists of studies on fuel performance and computer-based man-machine interfaces

Utilization of nuclear research reactors

International Nuclear Information System (INIS)

1980-01-01

prior to the beginning of the course was of particular value. Interesting scientific visits and demonstrations at the Isotope Institute and at the Central Research Institute for Physics (IFKI), both of the Hungarian Academy of Sciences, were also arranged. During the Study Tour at the Central Institute for Nuclear Research in Rossendorf near Dresden, German Democratic Republic, the participants had the opportunity to observe the organization of a 10 MW nuclear reactor where radioisotopes and radiopharmaceuticals are produced on a commercial scale. Lectures were delivered by local scientists on some of their programmes in applied research in solid state physics and material sciences. At the Technical University of Dresden, the group visited the homogeneous solid-moderated zero-power training reactor (AKR), primarily dedicated to nuclear education and training. Studies on different theoretical and experimental aspects of radiation protection (solid state nuclear track and thermoluminescent detectors) are also being carried out. The last day of the Study Tour was devoted to a visit to the College for Advanced Technology at Zittau, where a training reactor with a power of a few watts has been recently installed. (author)

Basic research and industrialization of CANDU advanced fuel

Energy Technology Data Exchange (ETDEWEB)

Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su [and others

2000-04-01

Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

Basic research and industrialization of CANDU advanced fuel

International Nuclear Information System (INIS)

Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su

2000-04-01

Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

Advanced gadolinia core and Toshiba advanced reactor management system

International Nuclear Information System (INIS)

Miyamoto, Toshiki; Yoshioka, Ritsuo; Ebisuya, Mitsuo

1988-01-01

At the Hamaoka Nuclear Power Station, Unit No. 3, advanced core design and core management technology have been adopted, significantly improving plant availability, operability and reliability. The outstanding technologies are the advanced gadolinia core (AGC) which utilizes gadolinium for the axial power distribution control, and Toshiba advanced reactor management system (TARMS) which uses a three-dimensional core physics simulator to calculate the power distribution. Presented here are the effects of these advanced technologies as observed during field testing. (author)

MARS: Mirror Advanced Reactor Study

International Nuclear Information System (INIS)

Logan, B.G.

1984-01-01

A recently completed two-year study of a commercial tandem mirror reactor design [Mirror Advanced Reactor Study (MARS)] is briefly reviewed. The end plugs are designed for trapped particle stability, MHD ballooning, balanced geodesic curvature, and small radial electric fields in the central cell. New technologies such as lithium-lead blankets, 24T hybrid coils, gridless direct converters and plasma halo vacuum pumps are highlighted

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

International Nuclear Information System (INIS)

1996-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-04-01

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

Present status of research reactor decommissioning programme in Indonesia

International Nuclear Information System (INIS)

Suripto, A.; Mulyanto, N.

2002-01-01

At present Indonesia has 3 research reactors, namely the 30 MW MTR-type multipurpose reactor at Serpong Site, two TRIGA-type research reactors, the first one being 1 MW located at Bandung Site and the second one a small reactor of 100 kW at Yogyakarta Site. The TRIGA Reactor at the Bandung Site reached its first criticality at 250 kW in 1964, and then was operated at 1000 kW since 1971. In October 2000 the reactor power was successfully upgraded to 2 MW. This reactor has already been operated for 38 years. There is not yet any decision for the decommissioning of this reactor. However it will surely be an object for the near future decommissioning programme and hence anticipation for the above situation becomes necessary. The regulation on decommissioning of research reactor is already issued by the independent regulatory body (BAPETEN) according to which the decommissioning permit has to be applied by the BATAN. For Indonesia, an early decommissioning strategy for research reactor dictates a restricted re-use of the site for other nuclear installation. This is based on high land price, limited availability of radwaste repository site, and other cost analysis. Spent graphite reflector from the Bandung TRIGA reactor is recommended for a direct disposal after conditioning, without any volume reduction treatment. Development of human resources, technological capability as well as information flow from and exchange with advanced countries are important factors for the future development of research reactor decommissioning programme in Indonesia. (author)

Global developments for advanced reactors and the role of the IAEA

International Nuclear Information System (INIS)

Kupitz, Juergen; Cleveland, John

1999-01-01

To assure that nuclear power can meet world energy needs in the near and medium term, considerable development activities are being carried out for each major reactor line, building on the large experience base. The programmes of global development activities for advanced nuclear power plants, and nuclear desalination are described. As an international forum for exchange of scientific and technical information, the IAEA plays a role in bringing together experts for a world-wide exchange of information about national programmes, trends in safety and user requirements, the impact of safety objectives on plant design, and the co-ordination of research programmes in advanced reactor technology. 15 refs

Development of demonstration advanced thermal reactor

Energy Technology Data Exchange (ETDEWEB)

Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige

1982-08-01

The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported.

Development of demonstration advanced thermal reactor

International Nuclear Information System (INIS)

Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige.

1982-01-01

The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported. (Kako, I.)

Comparison of advanced reactors program of different international vendors

International Nuclear Information System (INIS)

Agnihotri, N.K.

2001-01-01

The full text follows. Proposal for presenting a paper on Advanced Reactor Program Given below is the abstract for Track 6 session on Advanced Reactor at the ninth International Conference on Nuclear Engineering being held in Nice, France from April 8. through 12. 2001. This paper will provide an update on Advanced Reactor Program of different vendors in the United States, Japan, and Europe. Specifically the paper will look at the history of different Advanced Reactor Programs, international experience, aspect of economy due to standardization, and the highlights of technical specifications. The paper will also review aspects of Economy due to standardization, public acceptance, required construction time, and the experience of different vendors. The objective of the presentation is to underscore the highlights of the Reactor Program of different vendors in order to keep the attendees of the conference up-to-date. The presentation will be an impartial overview from an outsider's (not part of the Nuclear Steam Supply System's staff). (author)

New about research reactors

International Nuclear Information System (INIS)

Egorenkov, P.M.

2001-01-01

The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru

Summary of SMIRT20 Preconference Topical Workshop - Identifying Structural Issues in Advanced Reactors

International Nuclear Information System (INIS)

Richins, William; Novascone, Stephen; O'Brien, Cheryl

2009-01-01

The Idaho National Laboratory (INL, USA) and IASMiRT sponsored an international forum Nov 5-6, 2008 in Porvoo, Finland for nuclear industry, academic, and regulatory representatives to identify structural issues in current and future advanced reactor design, especially for extreme conditions and external threats. The purpose of this Topical Workshop was to articulate research, engineering, and regulatory Code development needs. The topics addressed by the Workshop were selected to address critical industry needs specific to advanced reactor structures that have long lead times and can be the subject of future SMiRT technical sessions. The topics were; (1) structural/materials needs for extreme conditions and external threats in contemporary (Gen. III) and future (Gen. IV and NGNP) advanced reactors and (2) calibrating simulation software and methods that address topic 1. The workshop discussions and research needs identified are presented. The Workshop successfully produced interactive discussion on the two topics resulting in a list of research and technology needs. It is recommended that IASMiRT communicate the results of the discussion to industry and researchers to encourage new ideas and projects. In addition, opportunities exist to retrieve research reports and information that currently exists, and encourage more international cooperation and collaboration. It is recommended that IASMiRT continue with an off-year workshop series on select topics.

Instrumentation for the advanced high-flux reactor workshop: proceedings

International Nuclear Information System (INIS)

Moon, R.M.

1984-01-01

The purpose of the Workshop on Instrumentation for the Advanced High-Flux Reactor, held on May 30, 1984, at the Oak Ridge National Laborattory, was two-fold: to announce to the scientific community that ORNL has begun a serious effort to design and construct the world's best research reactor, and to solicit help from the scientific community in planning the experimental facilities for this reactor. There were 93 participants at the workshop. We are grateful to the visiting scientists for their enthusiasm and interest in the reactor project. Our goal is to produce a reactor with a peak thermal flux in a large D 2 O reflector of 5 x 10 15 n/cm 2 s. This would allow the installation of unsurpassed facilities for neutron beam research. At the same time, the design will provide facilities for isotope production and materials irradiation which are significantly improved over those now available at ORNL. This workshop focussed on neutron beam facilities; the input from the isotope and materials irradiation communities will be solicited separately. The reactor project enjoys the full support of ORNL management; the present activities are financed by a grant of $663,000 from the Director's R and D Fund. However, we realize that the success of the project, both in realization and in use of the reactor, depends on the support and imagination of a broad segment of the scientific community. This is more a national project than an ORNL project. The reactor would be operated as a national user facility, open to any research proposal with high scientific merit. It is therefore important that we maintain a continuing dialogue with outside scientists who will be the eventual users of the reactor and the neutron beam facilities. The workshop was the first step in establishing this dialogue; we anticipate further workshops as the project continues

Windscale advanced gas-cooled reactor (WAGR) decommissioning project overview

International Nuclear Information System (INIS)

Pattinson, A.

2003-01-01

The current BNFL reactor decommissioning projects are presented. The projects concern power reactor sites at Berkely, Trawsfynydd, Hunterstone, Bradwell, Hinkley Point; UKAEA Windscale Pile 1; Research reactors within UK Scottish Universities at East Kilbride and ICI (both complete); WAGR. The BNFL environmental role include contract management; effective dismantling strategy development; implementation and operation; sentencing, encapsulation and transportation of waste. In addition for the own sites it includes strategy development; baseline decommissioning planning; site management and regulator interface. The project objectives for the Windscale Advanced Gas-Cooled Reactor (WAGR) are 1) Safe and efficient decommissioning; 2) Building of good relationships with customer; 3) Completion of reactor decommissioning in 2005. The completed WAGR decommissioning campaigns are: Operational Waste; Hot Box; Loop Tubes; Neutron Shield; Graphite Core and Restrain System; Thermal Shield. The current campaign is Lower Structures and the remaining are: Pressure vessel and Insulation; Thermal Columns and Outer Vault Membrane. An overview of each campaign is presented

Multipurpose research reactors

International Nuclear Information System (INIS)

1988-01-01

The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

Cost aspects of the research reactor fuel cycle

International Nuclear Information System (INIS)

2010-01-01

Research reactors have made valuable contributions to the development of nuclear power, basic science, materials development, radioisotope production for medicine and industry, and education and training. In doing so, they have provided an invaluable service to humanity. Research reactors are expected to make important contributions in the coming decades to further development of the peaceful uses of nuclear technology, in particular for advanced nuclear fission reactors and fuel cycles, fusion, high energy physics, basic research, materials science, nuclear medicine, and biological sciences. However, in the context of decreased public sector support, research reactors are increasingly faced with financial constraints. It is therefore of great importance that their operations are based on a sound understanding of the costs of the complete research reactor fuel cycle, and that they are managed according to sound financial and economic principles. This publication is targeted at individuals and organizations involved with research reactor operations, with the aim of providing both information and an analytical framework for assessing and determining the cost structure of fuel cycle related activities. Efficient management of fuel cycle expenditures is an important component in developing strategies for sustainable future operation of a research reactor. The elements of the fuel cycle are presented with a description of how they can affect the cost efficient operation of a research reactor. A systematic review of fuel cycle choices is particularly important when a new reactor is being planned or when an existing reactor is facing major changes in its fuel cycle structure, for example because of conversion of the core from high enriched uranium (HEU) to low enriched uranium (LEU) fuel, or the changes in spent fuel management provision. Review and optimization of fuel cycle issues is also recommended for existing research reactors, even in cases where research reactor

Research reactor support

International Nuclear Information System (INIS)

2005-01-01

Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

Series lecture on advanced fusion reactors

International Nuclear Information System (INIS)

Dawson, J.M.

1983-01-01

The problems concerning fusion reactors are presented and discussed in this series lecture. At first, the D-T tokamak is explained. The breeding of tritium and the radioactive property of tritium are discussed. The hybrid reactor is explained as an example of the direct use of neutrons. Some advanced fuel reactions are proposed. It is necessary to make physics consideration for burning advanced fuel in reactors. The rate of energy production and the energy loss are important things. The bremsstrahlung radiation and impurity radiation are explained. The simple estimation of the synchrotron radiation was performed. The numerical results were compared with a more detailed calculation of Taimor, and the agreement was quite good. The calculation of ion and electron temperature was made. The idea to use the energy more efficiently is that one can take X-ray or neutrons, and pass them through a first wall of a reactor into a second region where they heat the material. A method to convert high temperature into useful energy is the third problem of this lecture. The device was invented by A. Hertzberg. The lifetime of the reactor depends on the efficiency of energy recovery. The idea of using spin polarized nuclei has come up. The spin polarization gives a chance to achieve a large multiplication factor. The advanced fuel which looks easiest to make go is D plus He-3. The idea of multipole is presented to reduce the magnetic field inside plasma, and discussed. Two other topics are explained. (Kato, T.)

A Joint Report on PSA for New and Advanced Reactors

International Nuclear Information System (INIS)

2013-01-01

This report addresses the application of Probabilistic Safety Assessment (PSA) to new and advanced nuclear reactors. As far as advanced reactors are concerned, the objectives were to characterize the ability of current PSA technology to address key questions regarding the development, acceptance and licensing of advanced reactor designs, to characterize the potential value of advanced PSA methods and tools for application to advanced reactors, and to develop recommendations for any needed developments regarding PSA for these reactors. As far as the design and commissioning of new nuclear power plants is concerned, the objectives were to identify and characterize current practices regarding the role of PSA, to identify key technical issues regarding PSA, lessons learned and issues requiring further work; to develop recommendations regarding the use of PSA, and to identify future international cooperative work on the identified issues. In order to reach these objectives, questionnaires had been sent to participating countries and organisations

Safety classification of systems, structures, and components for pool-type research reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Ryong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2016-08-15

Structures, systems, and components (SSCs) important to safety of nuclear facilities shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions. Although SSC classification guidelines for nuclear power plants have been well established and applied, those for research reactors have been only recently established by the International Atomic Energy Agency (IAEA). Korea has operated a pool-type research reactor (the High Flux Advanced Neutron Application Reactor) and has recently exported another pool-type reactor (Jordan Research and Training Reactor), which is being built in Jordan. Korea also has a plan to build one more pool-type reactor, the Kijang Research Reactor, in Kijang, Busan. The safety classification of SSCs for pool-type research reactors is proposed in this paper based on the IAEA methodology. The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance. Because the SSCs in pool-type research reactors are not the pressure-retaining components, codes and standards for design of the SSCs following the safety classification can be selected in a graded approach.

A Design of Alarm System in a Research Reactor Facility

International Nuclear Information System (INIS)

Park, Jaekwan; Jang, Gwisook; Seo, Sangmun; Suh, Yongsuk

2013-01-01

The digital alarm system has become an indispensable design to process a large amount of alarms of power plants. Korean research reactor operated for decades maintains a hybrid alarm system with both an analog annunciator and a digital alarm display. In this design, several alarms are indicated on an analog panel and digital display, respectively, and it requires more attention and effort of the operators. As proven in power plants, a centralized alarm system design is necessary for a new research reactor. However, the number of alarms and operators in a research reactor is significantly lesser than power plants. Thus, simplification should be considered as an important factor for the operation efficiency. This paper introduces a simplified alarm system. As advances in information technology, fully digitalized alarm systems have been applied to power plants. In a new research reactor, it will be more useful than an analog or hybrid configuration installed in research reactors decades ago. However, the simplification feature should be considered as an important factor because the number of alarms and number of operators in a research reactor is significantly lesser than in power plants

Advanced Instrumentation for Transient Reactor Testing

Energy Technology Data Exchange (ETDEWEB)

Corradini, Michael L.; Anderson, Mark; Imel, George; Blue, Tom; Roberts, Jeremy; Davis, Kurt

2018-01-31

Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and design increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).

Advanced reactors: A retrospective

International Nuclear Information System (INIS)

Starr, C.

1989-01-01

The objectives for nuclear power have always emphasized competitive costs, reliability, and public safety. During its initial two decades, the nuclear reactor program was enthusiastically and generously supported by the public, government, and industry. In the subsequent decades this external support was substantially eroded by the growing public fears of catastrophic accidents, poor economic performance of many nuclear plants, regulatory constraints, and a plethora of engineering issues disclosed by plant operations. The technical and institutional histories are discussed with particular relevance to their influence on the framework for future development of the several proposed advance reactors

Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-04-20

The United States Department of Energy (DOE) commissioned the development of technology roadmaps for advanced (non-light water reactor) reactor concepts to help focus research and development funding over the next five years. The roadmaps show the research and development needed to support demonstration of an advanced (non-LWR) concept by the early 2030s, consistent with DOE’s Vision and Strategy for the Development and Deployment of Advanced Reactors. The intent is only to convey the technical steps that would be required to achieve such a goal; the means by which DOE will determine whether to invest in specific tasks will be treated separately. The starting point for the roadmaps is the Technical Readiness Assessment performed as part of an Advanced Test and Demonstration Reactor study released in 2016. The roadmaps were developed based upon a review of technical reports and vendor literature summarizing the technical maturity of each concept and the outstanding research and development needs. Critical path tasks for specific systems were highlighted on the basis of time and resources needed to complete the tasks and the importance of the system to the performance of the reactor concept. The roadmaps are generic, i.e. not specific to a particular vendor’s design but vendor design information may have been used as representative of the concept family. In the event that both near-term and more advanced versions of a concept are being developed, either a single roadmap with multiple branches or separate roadmaps for each version were developed. In each case, roadmaps point to a demonstration reactor (engineering or commercial) and show the activities that must be completed in parallel to support that demonstration in the 2030-2035 window. This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The SFR technology is mature enough for commercial demonstration by the early 2030s

Projecting regulatory expectations for advanced reactor designs

International Nuclear Information System (INIS)

Viktorov, A.

2011-01-01

This paper explores the overarching safety principles that will likely guide the safety design of advanced reactor technologies. As will be shown, the already established safety framework provides a solid foundation for the safety design of future nuclear power plants. As a specific example, the principle of 'proven technology' is presented in greater detail and its implications for a novel technology are discussed. Research, modeling and prototyping are shown to be components in satisfying this principle. While the fundamental safety principles are in place, their interpretation may depend both on the considered technology as well as the national context. Thus, the regulatory authority will need to be engaged, at an appropriate stage of the technology development, in specifying the regulatory requirements that will have to be met for a specific reactor design. (author)

Advances in Reactor Physics, Mathematics and Computation. Volume 1

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume one, are divided into 6 sessions bearing on: - session 1: Advances in computational methods including utilization of parallel processing and vectorization (7 conferences) - session 2: Fast, epithermal, reactor physics, calculation, versus measurements (9 conferences) - session 3: New fast and thermal reactor designs (9 conferences) - session 4: Thermal radiation and charged particles transport (7 conferences) - session 5: Super computers (7 conferences) - session 6: Thermal reactor design, validation and operating experience (8 conferences).

Metal fires and their implications for advanced reactors.

Energy Technology Data Exchange (ETDEWEB)

Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean; Hewson, John C.; Blanchat, Thomas K.

2010-10-01

This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in these areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety

Advanced spheromak fusion reactor

International Nuclear Information System (INIS)

Fowler, T.K.

1996-01-01

The spheromak has no toroidal magnetic field coils or other structure along its geometric axis, and is thus more attractive than the leading magnetic fusion reactor concept, the tokamak. As a consequence of this and other attributes, the spheromak reactor may be compact and produce a power density sufficiently high to warrant consideration of a liquid 'blanket' that breeds tritium, converts neutron kinetic energy to heat, and protects the reactor vessel from severe neutron damage. However, the physics is more complex, so that considerable research is required to learn how to achieve the reactor potential. Critical physics problems and possible ways of solving them are described. The opportunities and issues associated with a possible liquid wall are considered to direct future research

Overview on advanced nuclear reactors: research and deployment in the United States

International Nuclear Information System (INIS)

Sandell, L.; Rohrer, S.

2004-01-01

For the United States of America, the electricity requirement is expected to continue to rise at rates of approximately 1.8% over the next few years. This means that some 300,000 MW of additional generating capacity need to be made available by 2025. The Energy Policy Act of 2003 is to minimize this expected future growth of electricity consumption and promote research in favor of a diversified energy mix. As a consequence, the U.S. Senate and the House of Representatives passed legislation on electricity generation, on the promotion of, and research into, specific energy sources, and on energy conservation. Currently, coal-fired power plants contribute the largest share to the overall generating capacity. Considerable additions to the generating capacity have been made in the past ten years in gas-fired plants. In the light of the high present gas prices and market volatilities, the construction of new coal-fired power plants is currently under discussion. 103 out of the 436 nuclear power plants at present in operation worldwide are located in the United States. They represent by far the largest share of emission-free generating capacity in the United States. Considerable capacities have been added over the past few years by, up to now, 99 power increases by 0.4 to 17.8%. The Nuclear Power 2010 Program is a joint initiative by the government and industry seeking to further develop advanced nuclear power plant technologies and elaborate a new licensing procedure for nuclear power plants. The proposed licensing procedure and the Westinghouse AP1000, General Electric ESBWR, and AECL ACR-700 advanced reactor lines are presented. (orig.)

Issues affecting advanced passive light-water reactor safety analysis

International Nuclear Information System (INIS)

Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

1992-01-01

Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented

Evaluation of the trial design studies for an advanced marine reactor, (1)

International Nuclear Information System (INIS)

1988-03-01

The trial design of three type reactors, semi-integrated, integrated and integrated (self-pressurized) type, was carried out in order to clarify the reactor type for the advanced marine reactor that would be developed for its realization in future and in order to extract its research and development theme. The trial design was carried and finished as for the three type reactors in same specifications in order to improve the following characteristics, small in size, light in weight, high in safety and reliability, and economic. In this report, a comparison and review of the following items are described as for the above three type reactors, (1) specifications, (2) shielding, (3) refueling, (4) in-service inspection, (5) analysis of the transients and accidents, (6) piping systems, (7) control systems, (8) dynamic analysis, (9) overall comparison, (10) research and development theme and theme for study in future. (author)

The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

Activity report on the utilization of research reactors. Japanese fiscal year, 2004

International Nuclear Information System (INIS)

2006-03-01

During the fiscal year 2004, the Nuclear Science Research Institute research reactors carried out 7 cycles of joint use reactor operation at JRR-3 and 41 cycles at JRR-4. The research reactors are being utilized for various purposes including experimental studies such as neutron scattering, prompt gamma analysis, neutron radiography and medical irradiation (BNCT), and irradiation utilization such as neutron activation analysis of various samples, RI production, Irradiation Test of Reactor Materials and fission track, advanced Science Research. This volume contains 235 activity reports, which are categorized into the fields of neutron scattering (10 subcategories), neutron radiography, neutron activation analysis, RI Production, prompt gamma analysis, and others, submitted by the users in JAEA and from other organizations. (author)

Advanced Instrumentation, Information, and Control Systems Technologies Research in Support of Light Water Reactors

International Nuclear Information System (INIS)

Hallbert, Bruce P.; Kenneth, Thomas

2014-01-01

The Advanced Instrumentation, Information, and Control (II and C) Systems Technologies Pathway conducts targeted research and development (R and D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals to ensure that legacy analog II and C systems are not life-limiting issues for the LWR fleet, and to implement digital II and C technology in a manner that enables broad innovation and business improvement in the nuclear power plant operating model. Resolving long-term operational concerns with the II and C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation's energy and environmental security

Advanced Instrumentation, Information, and Control Systems Technologies Research in Support of Light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Hallbert, Bruce P.; Kenneth, Thomas [Idaho National Laboratory, Idaho (United States)

2014-08-15

The Advanced Instrumentation, Information, and Control (II and C) Systems Technologies Pathway conducts targeted research and development (R and D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals to ensure that legacy analog II and C systems are not life-limiting issues for the LWR fleet, and to implement digital II and C technology in a manner that enables broad innovation and business improvement in the nuclear power plant operating model. Resolving long-term operational concerns with the II and C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation's energy and environmental security.

Summary of advanced LMR [Liquid Metal Reactor] evaluations: PRISM [Power Reactor Inherently Safe Module] and SAFR [Sodium Advanced Fast Reactor

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Slovik, G.C.; Chan, B.C.; Kennett, R.J.; Cheng, H.S.; Kroeger, P.G.

1989-10-01

In support of the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) has performed independent analyses of two advanced Liquid Metal Reactor (LMR) concepts. The designs, sponsored by the US Department of Energy (DOE), the Power Reactor Inherently Safe Module (PRISM) [Berglund, 1987] and the Sodium Advanced Fast Reactor (SAFR) [Baumeister, 1987], were developed primarily by General Electric (GE) and Rockwell International (RI), respectively. Technical support was provided to DOE, RI, and GE, by the Argonne National Laboratory (ANL), particularly with respect to the characteristics of the metal fuels. There are several examples in both PRISM and SAFR where inherent or passive systems provide for a safe response to off-normal conditions. This is in contrast to the engineered safety systems utilized on current US Light Water Reactor (LWR) designs. One important design inherency in the LMRs is the ''inherent shutdown'', which refers to the tendency of the reactor to transition to a much lower power level whenever temperatures rise significantly. This type of behavior was demonstrated in a series of unscrammed tests at EBR-II [NED, 1986]. The second key design feature is the passive air cooling of the vessel to remove decay heat. These systems, designated RVACS in PRISM and RACS in SAFR, always operate and are believed to be able to prevent core damage in the event that no other means of heat removal is available. 27 refs., 78 figs., 3 tabs

International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

Energy Technology Data Exchange (ETDEWEB)

Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

2014-12-15

Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

Directions in advanced reactor technology

International Nuclear Information System (INIS)

Golay, M.W.

1990-01-01

Successful nuclear power plant concepts must simultaneously performance in terms of both safety and economics. To be attractive to both electric utility companies and the public, such plants must produce economical electric energy consistent with a level of safety which is acceptable to both the public and the plant owner. Programs for reactor development worldwide can be classified according to whether the reactor concept pursues improved safety or improved economic performance as the primary objective. When improved safety is the primary goal, safety enters the solution of the design problem as a constraint which restricts the set of allowed solutions. Conversely, when improved economic performance is the primary goal, it is allowed to be pursued only to an extent which is compatible with stringent safety requirements. The three major reactor coolants under consideration for future advanced reactor use are water, helium and sodium. Reactor development programs focuses upon safety and upon economics using each coolant are being pursued worldwide. These programs are discussed

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

Development of advanced boiling water reactor for medium capacity

International Nuclear Information System (INIS)

Kazuo Hisajima; Yutaka Asanuma

2005-01-01

This paper describes a result of development of an Advanced Boiling Water Reactor for medium capacity. 1000 MWe was selected as the reference. The features of the current Advanced Boiling Water Reactors, such as a Reactor Internal Pump, a Fine Motion Control Rod Drive, a Reinforced Concrete Containment Vessel, and three-divisionalized Emergency Core Cooling System are maintained. In addition, optimization for 1000 MWe has been investigated. Reduction in thermal power and application of the latest fuel reduced the number of fuel assemblies, Control Rods and Control Rod Drives, Reactor Internal Pumps, and Safety Relief Valves. The number of Main Steam lines was reduced from four to two. As for the engineered safety features, the Flammability Control System was removed. Special efforts were made to realize a compact Turbine Building, such as application of an in line Moisture Separator, reduction in the number of pumps in the Condensate and Feedwater System, and change from a Turbine-Driven Reactor Feedwater Pump to a Motor-Driven Reactor Feedwater Pump. 31% reduction in the volume of the Turbine Building is expected in comparison with the current Advanced Boiling Water Reactors. (authors)

Contributions of fast breeder test reactor to the advanced technology in India

International Nuclear Information System (INIS)

Kapoor, R.P.

2001-01-01

Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe loop type, sodium cooled, plutonium rich mixed carbide fuelled reactor. Its operation at Indira Gandhi Centre for Atomic Research, since first criticality in 1985, has contributed immensely to the advancement of this multidisciplinary and complex fast breeder technology in the country. It has also given a valuable operational feedback for the design of 500 MWe Prototype Fast Breeder Reactor. This paper highlights FBTR's significant contributions to this important technology which has a potential to provide energy security to the country in future. (author)

Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report

International Nuclear Information System (INIS)

Fromm, Bradley; Hauch, Benjamin; Sridharan, Kumar

2016-01-01

A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulation tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.

Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report

Energy Technology Data Exchange (ETDEWEB)

Fromm, Bradley [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hauch, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sridharan, Kumar [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-12-01

A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulation tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.

The in-core experimental program at the MIT Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Kohse, G.E.; Hu, L-W., E-mail: kohse@mit.edu [Massachusetts Inst. of Technology, Nuclear Reactor Lab., Cambridge, Massachusetts (United States)

2014-07-01

This paper describes the program of in-core experiments at the Massachusetts Institute of Technology Research Reactor (MITR), a 6 MW research reactor. The MITR has a neutron flux and spectrum similar to those in water-cooled power reactors and therefore provides a useful test environment for materials and fuels research. In-core facilities include: a water loop operating at pressurized water or boiling water reactor conditions, an inert gas irradiation facility operating at temperature up to 850 {sup o}C and special purpose facilities including fuel irradiation experiments. Recent and ongoing tests include: water loop investigations of corrosion and thermal and mechanical property evolution of SiC/SiC composites for fuel cladding, irradiation of advanced materials and in-core sensors at elevated temperatures, irradiation in molten fluoride salt at 700 {sup o}C of metal alloy, graphite and composite materials for power reactor applications and instrumented irradiations of metal-bonded hydride fuel. (author)

Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

Necessity of research reactors

International Nuclear Information System (INIS)

Ito, Tetsuo

2016-01-01

Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

Energy Technology Data Exchange (ETDEWEB)

Honma, George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-10-01

The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will be used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).

Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

International Nuclear Information System (INIS)

Honma, George

2015-01-01

The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will be used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).

Nuclear science and engineering education at a university research reactor

International Nuclear Information System (INIS)

Loveland, W.

1990-01-01

The research and teaching operations of the Nuclear Chemistry Division of the Dept. of Chemistry and the Dept. of Nuclear Engineering are housed at the Oregon State University Radiation Center. This facility which includes a 1.1 MW TRIGA reactor was used for 53 classes from a number of different academic departments last year. About one-half of these classes used the reactor and ∼25% of the reactor's 45 hour week was devoted to teaching. Descriptions will be given of reactor-oriented instructional programs in nuclear engineering, radiation health and nuclear chemistry. In nuclear chemistry, classes in (a) nuclear chemistry for nuclear engineers, (b) radiotracer methods, (c) elementary and advanced activation analysis, and (d) advanced nuclear instrumentation will be described in detail. The use of the facility to promote general nuclear literacy among college students, high school and grade school students and the general population will also be covered

Study on the Export Strategies for Research Reactors

International Nuclear Information System (INIS)

Oh, S. K.; Lee, Y. J.; Ham, T. K.; Hong, S. T.; Kim, J. H.

2008-12-01

Key strategic considerations taken into account should be based on understanding in the forecasts of demand and supply balance as well as the missions of research reactor for customers. For timely arrival at the competition, it may be advantageous to categorize the potential customers into 3 groups, the developed, the developing and the underdeveloped countries in respect of nuclear technology, and to be ready for the group-wise reference designs of the key reactor systems. Customizing the design to specific owner's requirements can advance from one of these reference designs when competition starts. To mobilize this approach effectively, it is useful to establish an integral project and technology management system earlier. This system will function as an important success factor for international research reactor business, because it makes easy to accommodate customer requirements and to achieve the design-to-cost.

Study on the Export Strategies for Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Oh, S. K.; Lee, Y. J.; Ham, T. K.; Hong, S. T.; Kim, J. H. [Ajou University, Suwon (Korea, Republic of)

2008-12-15

Key strategic considerations taken into account should be based on understanding in the forecasts of demand and supply balance as well as the missions of research reactor for customers. For timely arrival at the competition, it may be advantageous to categorize the potential customers into 3 groups, the developed, the developing and the underdeveloped countries in respect of nuclear technology, and to be ready for the group-wise reference designs of the key reactor systems. Customizing the design to specific owner's requirements can advance from one of these reference designs when competition starts. To mobilize this approach effectively, it is useful to establish an integral project and technology management system earlier. This system will function as an important success factor for international research reactor business, because it makes easy to accommodate customer requirements and to achieve the design-to-cost.

Reliability assurance for regulation of advanced reactors

International Nuclear Information System (INIS)

Fullwood, R.; Lofaro, R.; Samanta, P.

1992-01-01

The advanced nuclear power plants must achieve higher levels of safety than the first generation of plants. Showing that this is indeed true provides new challenges to reliability and risk assessment methods in the analysis of the designs employing passive and semi-passive protection. Reliability assurance of the advanced reactor systems is important for determining the safety of the design and for determining the plant operability. Safety is the primary concern, but operability is considered indicative of good and safe operation. this paper discusses several concerns for reliability assurance of the advanced design encompassing reliability determination, level of detail required in advanced reactor submittals, data for reliability assurance, systems interactions and common cause effects, passive component reliability, PRA-based configuration control system, and inspection, training, maintenance and test requirements. Suggested approaches are provided for addressing each of these topics

Reliability assurance for regulation of advanced reactors

International Nuclear Information System (INIS)

Fullwood, R.; Lofaro, R.; Samanta, P.

1991-01-01

The advanced nuclear power plants must achieve higher levels of safety than the first generation of plants. Showing that this is indeed true provides new challenges to reliability and risk assessment methods in the analysis of the designs employing passive and semi-passive protection. Reliability assurance of the advanced reactor systems is important for determining the safety of the design and for determining the plant operability. Safety is the primary concern, but operability is considered indicative of good and safe operation. This paper discusses several concerns for reliability assurance of the advanced design encompassing reliability determination, level of detail required in advanced reactor submittals, data for reliability assurance, systems interactions and common cause effects, passive component reliability, PRA-based configuration control system, and inspection, training, maintenance and test requirements. Suggested approaches are provided for addressing each of these topics

Multi purpose research reactor

International Nuclear Information System (INIS)

Raina, V.K.; Sasidharan, K.; Sengupta, Samiran; Singh, Tej

2006-01-01

At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor

Research and development into power reactor fuel performance

International Nuclear Information System (INIS)

Notley, M.J.F.

1983-07-01

The nuclear fuel in a power reactor must perform reliably during normal operation, and the consequences of abnormal events must be researched and assessed. The present highly reliable operation of the natural UO 2 in the CANDU power reactors has reduced the need for further work in this area; however a core of expertise must be retained for purposes such as training of new staff, retaining the capability of reacting to unforeseen circumstances, and participating in the commercial development of new ideas. The assessment of fuel performance during accidents requires research into many aspects of materials, fuel and fission product behaviour, and the consolidation of that knowledge into computer codes used to evaluate the consequences of any particular accident. This work is growing in scope, much is known from out-reactor work at temperatures up to about 1500 degreesC, but the need for in-reactor verification and investigation of higher-temperature accidents has necessitated the construction of a major new in-reactor test loop and the initiation of the associated out-reactor support programs. Since many of the programs on normal and accident-related performance are generic in nature, they will be applicable to advanced fuel cycles. Work will therefore be gradually transferred from the present, committed power reactor system to support the next generation of thorium-based reactor cycles

Operational experience of decommissioning techniques for research reactors in the United Kingdom

International Nuclear Information System (INIS)

England, M.R.; McCool, T.M.

2002-01-01

In previous co-ordinated research projects (CRP) conducted by the IAEA no distinction was made between decommissioning activities carried out at nuclear power plants, research reactors or nuclear fuel cycle facilities. As experience was gained and technology advanced it became clear that decommissioning of research reactors had certain specific characteristics which needed a dedicated approach. It was within this context that a CRP on Decommissioning Techniques for Research Reactors was launched and conducted by the IAEA from 1997 to 2001. This paper considers the experience gained from the decommissioning of two research reactors during the course of the CRP namely: (a) the ICI Triga Mk I reactor at Billingham UK which was largely complete by the end of the research project and (b) the Argonaut 100 reactor at the Scottish Universities Research and Reactor centre at East Kilbride in Scotland which is currently is the early stages of dismantling/site operations. It is the intention of this paper with reference to the two case studies outlined above to compare the actual implementation of these works against the original proposals and identify areas that were found to be problematical and/or identify any lessons learnt. (author)

Dynamic modeling of the advanced neutron source reactor

International Nuclear Information System (INIS)

March-Leuba, J.; Ibn-Khayat, M.

1990-01-01

The purpose of this paper is to provide a summary description and some applications of a computer model that has been developed to simulate the dynamic behavior of the advanced neutron source (ANS) reactor. The ANS dynamic model is coded in the advanced continuous simulation language (ACSL), and it represents the reactor core, vessel, primary cooling system, and secondary cooling systems. The use of a simple dynamic model in the early stages of the reactor design has proven very valuable not only in the development of the control and plant protection system but also of components such as pumps and heat exchangers that are usually sized based on steady-state calculations

Irradiation facilitates at the advanced test reactor

International Nuclear Information System (INIS)

Grover, Blaine S.

2006-01-01

The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC - formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950's with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens. The paper has the following contents: ATR description and capabilities; ATR operations, quality and safety requirements; Static capsule experiments; Lead experiments; Irradiation test vehicle; In-pile loop experiments; Gas test loop; Future testing; Support facilities at RTC; Conclusions. To summarize, the ATR has a long history in fuel and material irradiations, and will be fulfilling a critical role in the future fuel and material testing necessary to develop the next generation reactor systems and advanced fuel cycles. The

IAEA activities supporting the applications of research reactors in 2013

International Nuclear Information System (INIS)

Peld, N.D.; Ridikas, D.

2014-01-01

As the underutilization of research reactors around the world persists as a primary topic of concern among facility owners and operators, the IAEA responded in 2013 with a broad range of activities to address the planning, execution and improvement of many experimental techniques. The revision of two critical documents for planning and diversifying a facility's portfolio of applications, TECDOC 1234 'The Applications of Research Reactors' and TECDOC 1212 'Strategic Planning for Research Reactors', is in progress in order to keep this information relevant, corresponding to the dynamism of experimental techniques and research capabilities. Related to the latter TECDOC, the IAEA convened a meeting in 2013 for the expert review of a number of strategic plans submitted by research reactor operators in developing countries. A number of activities focusing on specific applications are either continuing or beginning as well. In neutron activation analysis, a joint round of inter-comparison proficiency testing sponsored by the IAEA Technical Cooperation Department will be completed, and facility progress in measurement accuracy is described. Also, a training workshop in neutron imaging and Coordinated Research Projects in reactor benchmarks, automation of neutron activation analysis and neutron beam techniques for material testing intend to advance these activities as more beneficial services to researchers and other users. (author)

Reactor safety research program. A description of current and planned reactor safety research sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research

International Nuclear Information System (INIS)

1975-06-01

The reactor safety research program, sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, is described in terms of its program objectives, current status, and future plans. Elements of safety research work applicable to water reactors, fast reactors, and gas cooled reactors are presented together with brief descriptions of current and planned test facilities. (U.S.)

Reactor physics innovations of the advanced CANDU reactor core: adaptable and efficient

International Nuclear Information System (INIS)

Chan, P.S.W.; Hopwood, J.M.; Bonechi, M.

2003-01-01

The Advanced CANDU Reactor (ACR) is designed to have a benign, operator-friendly core physics characteristic, including a slightly negative coolant-void reactivity and a moderately negative power coefficient. The discharge fuel burnup is about three times that of natural uranium fuel in current CANDU reactors. Key features of the reactor physics innovations in the ACR core include the use of H 2 O coolant, slightly enriched uranium (SEU) fuel, and D 2 O moderator in a reduced lattice pitch. These innovations result in substantial improvements in economics, as well as significant enhancements in reactor performance and waste reduction over the current reactor design. The ACR can be readily adapted to different power outputs by increasing or decreasing the number of fuel channels, while maintaining identical fuel and fuel-channel characteristics. The flexibility provided by on-power refuelling and simple fuel bundle design enables the ACR to easily adapt to the use of plutonium and thorium fuel cycles. No major modifications to the basic ACR design are required because the benign neutronic characteristics of the SEU fuel cycle are also inherent in these advanced fuel cycles. (author)

Advanced Carbothermal Electric Reactor, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — ORBITEC proposes to develop the Advanced Carbothermal Electric (ACE) reactor to efficiently extract oxygen from lunar regolith. Unlike state-of-the-art carbothermal...

Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

Energy Technology Data Exchange (ETDEWEB)

O' Brien, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yoon, Su -Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Housley, Gregory K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-09-01

This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

Research reactor DHRUVA

International Nuclear Information System (INIS)

Veeraraghaven, N.

1990-01-01

DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

CER. Research reactors in France

International Nuclear Information System (INIS)

Estrade, Jerome

2012-01-01

Networking and the establishment of coalitions between research reactors are important to guarantee a high technical quality of the facility, to assure well educated and trained personnel, to harmonize the codes of standards and the know-ledge of the personnel as well as to enhance research reactor utilization. In addition to the European co-operation, country-specific working groups have been established for many years, such as the French research reactor Club d'Exploitants des Reacteurs (CER). It is the association of French research reactors representing all types of research reactors from zero power up to high flux reactors. CER was founded in 1990 and today a number of 14 research reactors meet twice a year for an exchange of experience. (orig.)

Applications of Research Reactors

International Nuclear Information System (INIS)

2014-01-01

One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world.' One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property'. The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. The purpose of the earlier publication, The Application of Research Reactors, IAEA-TECDOC-1234, was to present descriptions of the typical forms of research reactor use. The necessary criteria to enable an application to be performed were outlined for each one, and, in many cases, the minimum as well as the desirable requirements were given. This revision of the publication over a decade later maintains the original purpose and now specifically takes into account the changes in service requirements demanded by the relevant stakeholders. In particular, the significant improvements in

Assessment of core protection and monitoring systems for an advanced reactor SMART

International Nuclear Information System (INIS)

In, Wang Kee; Hwang, Dae Hyun; Yoo, Yeon Jong; Zee, Sung Qunn

2002-01-01

Analogue and digital core protection/monitoring systems were assessed for the implementation in an advanced reactor. The core thermal margins to nuclear fuel design limits (departure from nucleate boiling and fuel centerline melting) were estimated using the design data for a commercial pressurized water reactor and an advanced reactor. The digital protection system resulted in a greater power margin to the fuel centerline melting by at least 30% of rated power for both commercial and advanced reactors. The DNB margin with the digital system is also higher than that for the analogue system by 8 and 12.1% of rated power for commercial and advanced reactors, respectively. The margin gain with the digital system is largely due to the on-line calculations of DNB ratio and peak local power density from the live sensor signals. The digital core protection and monitoring systems are, therefore, believed to be more appropriate for the advanced reactor

INEEL Advanced Radiotherapy Research Program Annual Report 2001

International Nuclear Information System (INIS)

Venhuizen, James R.

2002-01-01

This report summarizes the major activities and accomplishments of the Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Radiotherapy Research Program for calendar year 2001. Applications of supportive research and development, as well as technology deployment in the fields of chemistry, radiation physics and dosimetry, and neutron source design and demonstration are described. Contributions in the fields of physics and biophysics include development of advanced patient treatment planning software, feasibility studies of accelerator neutron source technology for Neutron Capture Therapy (NCT), and completion of major modifications to the research reactor at Washington State University to produce an epithermal-neutron beam for NCT research applications

INEEL Advanced Radiotherapy Research Program Annual Report 2001

Energy Technology Data Exchange (ETDEWEB)

Venhuizen, James Robert

2002-04-01

This report summarizes the major activities and accomplishments of the Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Radiotherapy Research Program for calendar year 2001. Applications of supportive research and development, as well as technology deployment in the fields of chemistry, radiation physics and dosimetry, and neutron source design and demonstration are described. Contributions in the fields of physics and biophysics include development of advanced patient treatment planning software, feasibility studies of accelerator neutron source technology for Neutron Capture Therapy (NCT), and completion of major modifications to the research reactor at Washington State University to produce an epithermal-neutron beam for NCT research applications.

INEEL Advanced Radiotherapy Research Program Annual Report 2001

Energy Technology Data Exchange (ETDEWEB)

Venhuizen, James R.

2002-04-30

This report summarizes the major activities and accomplishments of the Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Radiotherapy Research Program for calendar year 2001. Applications of supportive research and development, as well as technology deployment in the fields of chemistry, radiation physics and dosimetry, and neutron source design and demonstration are described. Contributions in the fields of physics and biophysics include development of advanced patient treatment planning software, feasibility studies of accelerator neutron source technology for Neutron Capture Therapy (NCT), and completion of major modifications to the research reactor at Washington State University to produce an epithermal-neutron beam for NCT research applications.

Research Reactors for the Development of Materials and Fuels for Innovative Nuclear Energy Systems

International Nuclear Information System (INIS)

2017-01-01

This publication presents an overview of research reactor capabilities and capacities in the development of fuels and materials for innovative nuclear reactors, such as GenIV reactors. The compendium provides comprehensive information on the potential for materials and fuel testing research of 30 research reactors, both operational and in development. This information includes their power levels, mode of operation, current status, availability and historical overview of their utilization. A summary of these capabilities and capacities is presented in the overview tables of section 6. Papers providing a technical description of the research reactors, including their specific features for utilization are collected as profiles on a CD-ROM and represent an integral part of this publication. The publication is intended to foster wider access to information on existing research reactors with capacity for advanced material testing research and thus ensure their increased utilization in this particular domain. It is expected that it can also serve as a supporting tool for the establishment of regional and international networking through research reactor coalitions and IAEA designated international centres based on research reactors.

Advanced reactors and future energy market needs

International Nuclear Information System (INIS)

Paillere, Henri; )

2017-01-01

Based on the results of a very well-attended international workshop on 'Advanced Reactor Systems and Future Energy Market Needs' that took place in April 2017, the NEA has embarked on a two-year study with the objective of analysing evolving energy market needs and requirements, as well as examining how well reactor technologies under development today will fit into tomorrow's low-carbon world. The NEA Expert Group on Advanced Reactor Systems and Future Energy Market Needs (ARFEM) held its first meeting on 5-6 July 2017 with experts from Canada, France, Italy, Japan, Korea, Poland, Romania, Russia and the United Kingdom. The outcome of the study will provide much needed insight into how well nuclear can fulfil its role as a key low-carbon technology, and help identify challenges related to new operational, regulatory or market requirements

Updated comparison of economics of fusion reactors with advanced fission reactors

International Nuclear Information System (INIS)

Delene, J.G.

1990-01-01

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

Computerized reactor monitor and control for research reactors

International Nuclear Information System (INIS)

Buerger, L.; Vegh, E.

1981-09-01

The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

Nuclear reactor instrumentation at research reactor renewal

International Nuclear Information System (INIS)

Baers, B.; Pellionisz, P.

1981-10-01

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

Strategic planning for research reactors. Guidance for reactor managers

International Nuclear Information System (INIS)

2001-04-01

The purpose of this publication is to provide guidance on how to develop a strategic plan for a research reactor. The IAEA is convinced of the need for research reactors to have strategic plans and is issuing a series of publications to help owners and operators in this regard. One of these covers the applications of research reactors. That report brings together all of the current uses of research reactors and enables a reactor owner or operator to evaluate which applications might be possible with a particular facility. An analysis of research reactor capabilities is an early phase in the strategic planning process. The current document provides the rationale for a strategic plan, outlines the methodology of developing such a plan and then gives a model that may be followed. While there are many purposes for research reactor strategic plans, this report emphasizes the use of strategic planning in order to increase utilization. A number of examples are given in order to clearly illustrate this function

Development of advanced nuclear core analysis system applicable to various reactor types

International Nuclear Information System (INIS)

Kaneko, Kunio

2002-03-01

This fiscal year, aiming at development of an advanced detailed analysis system applicable to nuclear core performance analysis of various fast reactors currently considered, the concept of cross section library set was examined and the specification of library set was determined. That is to say, referring the world most advanced reactor physics analysis system ERANOS (European Reactor Analysis Optimized System) and the result of preceding research 'preparation of next generation cross section library', 900 energy groups structure, concrete cross section data to be included and the format of cross section library were defined. And we performed elaborate work revising the group cross section production system which was prepared in the preceding research. After that the revision work was completed, to confirm the capability of revised cross section production system, we produced a prototype 450 groups cross section library. And we carried out a series of bench mark tests including analysis of small fast reactors utilizing this prototype cross section library and confirmed that the prototype cross section library has sufficient accuracy for predicting core performance. Furthermore, we estimated the computer resource information such as memory size, hard disk capacity and calculation time, etc. necessary for producing 900 groups detailed cross section library. In addition, we identified problems to be solved for developing a cell calculation code installed in our detailed analysis system. (author)

Advances in U.S. reactor physics standards

International Nuclear Information System (INIS)

Cokinos, Dimitrios

2008-01-01

The standards for Reactor Design, widely used in the nuclear industry, provide guidance and criteria for performing and validating a wide range of nuclear reactor calculations and measurements. Advances, over the past decades in reactor technology, nuclear data and infrastructure in the data handling field, led to major improvements in the development and application of reactor physics standards. A wide variety of reactor physics methods and techniques are being used by reactor physicists for the design and analysis of modern reactors. ANSI (American National Standards Institute) reactor physics standards, covering such areas as nuclear data, reactor design, startup testing, decay heat and fast neutron fluence in the pressure vessel, are summarized and discussed. These standards are regularly undergoing review to respond to an evolving nuclear technology and are being successfully used in the U.S and abroad contributing to improvements in reactor design, safe operation and quality assurance. An overview of the overall program of reactor physics standards is presented. New standards currently under development are also discussed. (authors)

Ageing of research reactors

International Nuclear Information System (INIS)

Ciocanescu, M.

2001-01-01

Historically, many of the research institutions were centred on a research reactor facility as main technological asset and major source of neutrons for research. Important achievements were made in time in these research institutions for development of nuclear materials technology and nuclear safety for nuclear energy. At present, ageing of nuclear research facilities among these research reactors and ageing of staff are considerable factors of reduction of competence in research centres. The safe way of mitigation of this trend deals with ageing management by so called, for power reactors, Plant Life Management and new investments in staff as investments in research, or in future resources of competence. A programmatic approach of ageing of research reactors in correlation with their actual and future utilisation, will be used as a basis for safety evaluation and future spending. (author)

Advanced Test Reactor probabilistic risk assessment

International Nuclear Information System (INIS)

Atkinson, S.A.; Eide, S.A.; Khericha, S.T.; Thatcher, T.A.

1993-01-01

This report discusses Level 1 probabilistic risk assessment (PRA) incorporating a full-scope external events analysis which has been completed for the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory

Siting of research reactors

International Nuclear Information System (INIS)

1987-01-01

The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

International Nuclear Information System (INIS)

Forsberg, C.W.

1989-01-01

Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' [Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety] is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document

Digital control of research reactors

International Nuclear Information System (INIS)

Crump, J.C. III.; Richards, W.J.; Heidel, C.C.

1991-01-01

Research reactors provide an important service for the nuclear industry. Developments and innovations used for research reactors can be later applied to larger power reactors. Their relatively inexpensive cost allows research reactors to be an excellent testing ground for the reactors of tomorrow. One area of current interest is digital control of research reactor systems. Digital control systems offer the benefits of implementation and superior system response over their analog counterparts. At McClellan Air Force Base in Sacramento, California, the Stationary Neutron Radiography System (SNRS) uses a 1,000-kW TRIGA reactor for neutron radiography and other nuclear research missions. The neutron radiography beams generated by the reactor are used to detect corrosion in aircraft structures. While the use of the reactor to inspect intact F-111 wings is in itself noteworthy, there is another area in which the facility has applied new technology: the instrumentation and control system (ICS). The ICS developed by General Atomics (GA) contains several new and significant items: (a) the ability to servocontrol on three rods, (b) the ability to produce a square wave, and (c) the use of a software configurator to tune parameters affected by the actual reactor core dynamics. These items will probably be present in most, if not all, future research reactors. They were developed with increased control and overall usefulness of the reactor in mind

3. Research Coordination Meeting (RCM) of the Coordinated Research Project (CRP) on 'Studies of advanced reactor technology options for effective incineration of radioactive waste'. Working material

International Nuclear Information System (INIS)

2007-01-01

To meet expressed Member States' needs, the IAEA has initiated a Coordinated Research Project (CRP) on 'Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste'. The final goal of the CRP is to deepen the understanding of the dynamics of transmutation systems, e.g. the accelerator driven system, especially systems with deteriorated safety parameters, qualify the available methods, specify the range of validity of methods, and formulate requirements for future theoretical developments. Should transient experiments be available, the CRP will pursue experimental benchmarking work. In any case, based on the results, the CRP will conclude on the potential need of transient experiments and make appropriate proposals for experimental programs. The Technical Meeting in Chennai was the 3rd Research Coordination Meeting (RCM) of the CRP The man objectives of the RCM were to: - Discuss and perform inter-comparisons of the various benchmark results; - Prepare the first draft of the final CRP Report Status of the analyses and inter-comparisons of the results. The main objective of the CRP was to study innovative technology options for incinerating/utilizing radioactive wastes. The CRP's benchmarking exercises focused on eight innovative transmutation 'Domains', which correspond to different critical and sub-critical concepts or groups of concepts: I. Critical fast reactor, solid fuel, with fertile; II. Critical fast reactor, solid fuel, fertile-free; III. ADS, solid fuel, with fertile; IV. ADS, solid fuel, fertile-free; V. Critical reactor and ADS, molten salt fuel, with fertile; VI. Critical reactor and ADS, molten salt fuel, fertile-free; VII. Critical fast reactor and ADS, gas cooled; VIII. Fusion/fission hybrid system. For each of these Domains, the discussions and inter-comparisons considered the following issues: - Reactor-models; - Scenarios/phenomena; - Static analyses; - Dynamic analyses; - Methods; - Codes; - Neutronic data base

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-01

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

State of the art of the advanced pressurized water reactor

International Nuclear Information System (INIS)

Seifritz, W.; Chawla, R.

1987-01-01

A review is given of the present status of the works concerned with an advanced pressurized water reactor (APWR). It includes the following items: reactor physics, thermal and hydraulic investigations and other engineering aspects as well as an analysis of electricity generation cost and long-term problems of embedding the APWR in a plutonium economy. As a summary it can be stated that there are discernible no principal obstacles of technically accomplishing an APWR, but there will be necessary considerable expenses in research and development works if it should be intended to start commercial service of an APWR up to the end of this century. (author)

IAEA safety standards for research reactors

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

Research reactors and materials testing

International Nuclear Information System (INIS)

Vidal, H.

1986-01-01

Research reactors can be classified in three main groups according to the moderator which is used. Their technical characteristics are given and the three most recent research and materials testing reactors are described: OSIRIS, ORPHEE and the high-flux reactor of Grenoble. The utilization of research reactors is reviewed in four fields of activity: training, fundamental or applied research and production (eg. radioisotopes) [fr

Directory of Nuclear Research Reactors 1994

International Nuclear Information System (INIS)

1995-08-01

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

Training reactor deployment. Advanced experimental course on designing new reactor cores

International Nuclear Information System (INIS)

Skoda, Radek

2009-01-01

Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

Physical security at research reactors

International Nuclear Information System (INIS)

Clark, R.A.

1977-01-01

Of the 84 non-power research facilities licensed under 10 CFR Part 50, 73 are active (two test reactors, 68 research reactors and three critical facilities) and are required by 10 CFR Part 73.40 to provide physical protection against theft of SNM and against industrial sabotage. Each licensee has developed a security plan required by 10 CFR Part 50.34(c) to demonstrate the means of compliance with the applicable requirements of 10 CFR Part 73. In 1974, the Commission provided interim guidance for the organization and content of security plans for (a) test reactors, (b) medium power research and training reactors, and (c) low power research and training reactors. Eleven TRIGA reactors, with power levels greater than 250 kW and all other research and training reactors with power levels greater than 100 kW and less than or equal to 5,000 kW are designated as medium power research and training reactors. Thirteen TRIGA reactors with authorized power levels less than 250 kW are considered to be low power research and training reactors. Additional guidance for complying with the requirements of 73.50 and 73.60, if applicable, is provided in the Commission's Regulatory Guides. The Commission's Office of Inspection and Enforcement inspects each licensed facility to assure that an approved security plan is properly implemented with appropriate procedures and physical protection systems

TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

International Nuclear Information System (INIS)

Grover, S.B.

2004-01-01

The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor

International Nuclear Information System (INIS)

S. Blaine Grover

2004-01-01

The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

An advanced liquid hydrogen cold source for the NIST research reactor

International Nuclear Information System (INIS)

Williams, R.E.; Kopetka, P.; Rowe, J.M.

1999-01-01

A second-generation liquid hydrogen cold neutron source is currently being fabricated and will be installed in the NIST reactor early next year. The existing source has operated very successfully over the last four years, providing a six-fold increase in the cold neutron yield compared to the previous heavy ice source. The design of the new source is based on our operating experience with the existing LH 2 source and extensive neutron transport calculations using improved MCNP modeling and computational capabilities. Enhanced mechanical design and manufacturing tools are exploited in the fabrication of the advanced source, which is expected to nearly double the yield of the existing LH 2 source. (author)

Management of research reactor ageing

International Nuclear Information System (INIS)

1995-03-01

As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs

Management of research reactor ageing

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-03-01

As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs.

Material and component progress within ARCHER for advanced high temperature reactor

International Nuclear Information System (INIS)

Buckthorpe, D.E.; Davies, M.; Pra, F.; Bonnamy, P.; Fokkens, J.; Heijna, M.; Bout, N. de; Vreeling, A.; Bourlier, F.; Lhachemi, D.; Woayehune, A.; Dubiez-le-Goff, S.; Hahner, P.; Futterer, M.; Berka, J.; Kalivodora, J.; Pouchon, M.A.; Schmitt, R.; Homerin, P.; Marsden, B.; Mummery, P.; Mutch, G.; Ponca, D.; Buhl, P.; Hoffmann, M.; Rondet, F.; Pecherty, A.; Baurand, F.; Alenda, F.; Esch, M.; Kohlz, N.; Reed, J.; Fachinger, J.; Klower, Dr.

2014-01-01

The ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R and D) integrated project started in 2011 as part of the European Commission 7. Framework Programme (FP7) for a period of four years to perform High Temperature Reactor technology R and D in support of reactor demonstration. The project consortium encompasses conventional and Nuclear Industry, Utilities, Technical Support Organizations, Research and Development Organizations and Academia. The activities involved contribute to the Generation IV (GIF) International Forum and collaborate with related projects in the US, China, Japan, and the Republic of Korea in cooperation with IAEA and ISTC. This paper addresses the progress of the work on materials and component technologies within ARCHER over the first two years of the project. (authors)

Advanced Reactor Technology/Energy Conversion Project FY17 Accomplishments.

Energy Technology Data Exchange (ETDEWEB)

Rochau, Gary E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2018-02-01

The purpose of the ART Energy Conversion (EC) Project is to provide solutions to convert the heat from an advanced reactor to useful products that support commercial application of the reactor designs.

Status of the advanced boiling water reactor and simplified boiling water reactor

International Nuclear Information System (INIS)

Smith, P.F.

1992-01-01

This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

Research and materials irradiation reactors

International Nuclear Information System (INIS)

Ballagny, A.; Guigon, B.

2004-01-01

Devoted to the fundamental and applied research on materials irradiation, research reactors are nuclear installations where high neutrons flux are maintained. After a general presentation of the research reactors in the world and more specifically in France, this document presents the heavy water cooled reactors and the water cooled reactors. The third part explains the technical characteristics, thermal power, neutron flux, operating and details the Osiris, the RHF (high flux reactor), the Orphee and the Jules Horowitz reactors. The last part deals with the possible utilizations. (A.L.B.)

Advances in High Temperature Gas Cooled Reactor Fuel Technology

International Nuclear Information System (INIS)

2012-12-01

This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

Advances in High Temperature Gas Cooled Reactor Fuel Technology

International Nuclear Information System (INIS)

2012-06-01

This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

Filling the gaps in SCWR materials research: advanced nuclear corrosion research facilities in Hamilton

International Nuclear Information System (INIS)

Krausher, J.L.; Zheng, W.; Li, J.; Guzonas, D.; Botton, G.

2011-01-01

Research efforts on materials selection and development in support of the design of supercritical water-cooled reactors (SCWRs) have produced a considerable amount of data on corrosion, creep and other related properties. Summaries of the data on corrosion [1] and stress corrosion cracking [2] have recently been produced. As research on the SCWR advances, gaps and limitations in the published data are being identified. In terms of corrosion properties, these gaps can be seen in several areas, including: 1) the test environment, 2) the physical and chemical severity of the tests conducted as compared with likely reactor service/operating conditions, and 3) the test methods used. While some of these gaps can be filled readily using existing facilities, others require the availability of advanced test facilities for specific tests and assessments. In this paper, highlights of the new materials research facilities jointly established in Hamilton by CANMET Materials Technology Laboratory and McMaster University are presented. (author)

Reactor Materials Research

International Nuclear Information System (INIS)

Van Walle, E.

2001-01-01

The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

Reactor Materials Research

Energy Technology Data Exchange (ETDEWEB)

Van Walle, E

2001-04-01

The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

Energy Technology Data Exchange (ETDEWEB)

1993-05-13

Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

The research reactors their contribution to the reactors physics

International Nuclear Information System (INIS)

Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

2000-01-01

The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

Conjugate heat transfer simulations of advanced research reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Piro, M.H.A., E-mail: pirom@aecl.ca; Leitch, B.W.

2014-07-01

Highlights: • Temperature predictions are enhanced by coupling heat transfer in solid and fluid zones. • Seven different cases are considered to observe trends in predicted temperature and pressure. • The seven cases consider high/medium/low power, flow, burnup, fuel material and geometry. • Simulations provide temperature predictions for performance/safety. Boiling is unlikely. • Simulations demonstrate that a candidate geometry can enhance performance/safety. - Abstract: The current work presents numerical simulations of coupled fluid flow and heat transfer of advanced U–Mo/Al and U–Mo/Mg research reactor fuels in support of performance and safety analyses. The objective of this study is to enhance predictions of the flow regime and fuel temperatures through high fidelity simulations that better capture various heat transfer pathways and with a more realistic geometric representation of the fuel assembly in comparison to previous efforts. Specifically, thermal conduction, convection and radiation mechanisms are conjugated between the solid and fluid regions. Also, a complete fuel element assembly is represented in three dimensional space, permitting fluid flow and heat transfer to be simulated across the entire domain. Seven case studies are examined that vary the coolant inlet conditions, specific power, and burnup to investigate the predicted changes in the pressure drop in the coolant and the fuel, clad and coolant temperatures. In addition, an alternate fuel geometry is considered with helical fins (replacing straight fins in the existing design) to investigate the relative changes in predicted fluid and solid temperatures. Numerical simulations predict that the clad temperature is sensitive to changes in the thermal boundary layer in the coolant, particularly in simultaneously developing flow regions, while the temperature in the fuel is anticipated to be unaffected. Finally, heat transfer between fluid and solid regions is enhanced with

Advanced test reactor testing experience-past, present and future

International Nuclear Information System (INIS)

Marshall, Frances M.

2006-01-01

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans

Global trends in advanced reactor developments, and the role of the IAEA

International Nuclear Information System (INIS)

Kupitz, J.; Cleveland, J.

2000-01-01

In the second half of the 20 t 'h century nuclear power has evolved from the research and development environment to an industry that supplies approximately 16% of the world's electricity. In these 50 years of nuclear development a great deal has been achieved and many lessons have been learned. At the end of 1998, according to data reported in the Power Reactor Information System, PRlS, of the IAEA, there were 434 nuclear power plants in operation and 36 under construction. About 9500 reactor-years of operating experience have been accumulated by today. World-wide, considerable efforts are being made to develop advanced nuclear power. Various organizations are involved, including governments, industries, utilities, universities, national laboratories, and research institutes. Expenditures for development of new designs, technology improvements, and the related research for the major reactor types combined is estimated to exceed US$ 2 billion per year. This paper gives an overview about nuclear power technology development programmes and projects in Member States and the role of the IAEA as a forum for informatics exchange and co-operative research. (authors)

Next generation advanced nuclear reactor designs

International Nuclear Information System (INIS)

Turgut, M. H.

2009-01-01

Growing energy demand by technological developments and the increase of the world population and gradually diminishing energy resources made nuclear power an indispensable option. The renewable energy sources like solar, wind and geothermal may be suited to meet some local needs. Environment friendly nuclear energy which is a suitable solution to large scale demands tends to develop highly economical, advanced next generation reactors by incorporating technological developments and years of operating experience. The enhancement of safety and reliability, facilitation of maintainability, impeccable compatibility with the environment are the goals of the new generation reactors. The protection of the investment and property is considered as well as the protection of the environment and mankind. They became economically attractive compared to fossil-fired units by the use of standard designs, replacing some active systems by passive, reducing construction time and increasing the operation lifetime. The evolutionary designs were introduced at first by ameliorating the conventional plants, than revolutionary systems which are denoted as generation IV were verged to meet future needs. The investigations on the advanced, proliferation resistant fuel cycle technologies were initiated to minimize the radioactive waste burden by using new generation fast reactors and ADS transmuters.

Improvement of research reactor sustainability

International Nuclear Information System (INIS)

Ciocanescu, M.; Paunoiu, C.; Toma, C.; Preda, M.; Ionila, M.

2010-01-01

The Research Reactors as is well known have numerous applications in a wide range of science technology, nuclear power development, medicine, to enumerate only the most important. The requirements of clients and stack-holders are fluctuating for the reasons out of control of Research Reactor Operating Organization, which may ensure with priority the safety of facility and nuclear installation. Sustainability of Research Reactor encompasses several aspects which finally are concentrated on safety of Research Reactor and economical aspects concerning operational expenses and income from external resources. Ensuring sustainability is a continuous, permanent activity and also it requests a strategic approach. The TRIGA - 14 MW Research Reactor detains a 30 years experience of safe utilization with good performance indicators. In the last 4 years the reactor benefited of a large investment project for modernization, thus ensuring the previous performances and opening new perspectives for power increase and for new applications. The previous core conversion from LEU to HEU fuel accomplished in 2006 ensures the utilization of reactor based on new qualified European supplier of TRIGA LEU fuel. Due to reduction of number of performed research reactors, the 14 MW TRIGA modernized reactor will play a significant role for the following two decades. (author)

Advanced Small Modular Reactor Economics Status Report

Energy Technology Data Exchange (ETDEWEB)

Harrison, Thomas J [ORNL

2014-10-01

This report describes the data collection work performed for an advanced small modular reactor (AdvSMR) economics analysis activity at the Oak Ridge National Laboratory. The methodology development and analytical results are described in separate, stand-alone documents as listed in the references. The economics analysis effort for the AdvSMR program combines the technical and fuel cycle aspects of advanced (non-light water reactor [LWR]) reactors with the market and production aspects of SMRs. This requires the collection, analysis, and synthesis of multiple unrelated and potentially high-uncertainty data sets from a wide range of data sources. Further, the nature of both economic and nuclear technology analysis requires at least a minor attempt at prediction and prognostication, and the far-term horizon for deployment of advanced nuclear systems introduces more uncertainty. Energy market uncertainty, especially the electricity market, is the result of the integration of commodity prices, demand fluctuation, and generation competition, as easily seen in deregulated markets. Depending on current or projected values for any of these factors, the economic attractiveness of any power plant construction project can change yearly or quarterly. For long-lead construction projects such as nuclear power plants, this uncertainty generates an implied and inherent risk for potential nuclear power plant owners and operators. The uncertainty in nuclear reactor and fuel cycle costs is in some respects better understood and quantified than the energy market uncertainty. The LWR-based fuel cycle has a long commercial history to use as its basis for cost estimation, and the current activities in LWR construction provide a reliable baseline for estimates for similar efforts. However, for advanced systems, the estimates and their associated uncertainties are based on forward-looking assumptions for performance after the system has been built and has achieved commercial operation

Results on safety research for five years (from fiscal year 1996 to 2000). A field of power reactors

International Nuclear Information System (INIS)

2001-10-01

This safety research carried out by the Japan Nuclear Cycle Development Institute (JNC) for five years ranged from 1996 to 2000 fiscal year, was performed according to the safety research basic plan (from 1996 to 2000 fiscal year) established on March, 1996 (revised again on May, 2000). This report was arranged on a field on power reactors (all subjects on fields of advanced conversion reactor and a subject on power reactor in a field of seismic resistant and probability theoretical safety evaluation) by combining its research results for five years ranged from 1996 to 2000 fiscal year with general outlines on the safety research basic plan. Here were shown outlines on the safety research basic plan, aims and subjects on safety research at a field of power reactors, a list of survey sheets on safety research result, and survey sheets on safety research results. The survey sheets containing research field, title, organization, researcher name, researching period, names of cooperative organization, using facilities, research outline, research results, established contents, application, and research trends, are ranged to 5 items on advanced conversion reactor, 29 items on high breeder reactor, 1 item on seismic resistance, and 5 items on probability theoretical safety evaluation. (G.K.)

National nuclear power planning of China and advanced reactor

International Nuclear Information System (INIS)

Qian Jihui

1990-01-01

The necessity of investigation on the trends of advanced reactor technology all over the world is elabrated while China is going to set up its long-term national nuclear power programme. In author's opinion, thermal reactor power plants will have a quite long period development in the next century and a new trend of second generation NPPs might emerge in the beginning of next century. These new generation advanced reactors are characterized with new design concepts based on the inherent or passive safety features. Among them, most promising ones are those of AP-600 and MHTGR. Chinese experts are paying special attention to and closely following these two directions

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

Workshop on PSA for New and Advanced Reactors

International Nuclear Information System (INIS)

2012-01-01

This workshop was organized by the NEA Working Group on Risk Assessment (WGRISK). The key objective of the workshop was to share the current state-of-the art on the PSA (Probabilistic Safety Assessment) applied for new reactors and advanced reactors. Fifty experts from 13 countries and one international organization (IAEA) participated in the present workshop, and 35 technical papers were presented. The main topics of interest, discussed during the workshop, included the followings: regulatory aspects, risk-informed methods, technical aspects of the PSA for new and advanced reactors, hazards of PSA (internal and external), severe accident/source term/Level 2 PSA, and consequence analysis/Level 3 PSA. Among the technical aspects of the PSA, the assessment of the reliability of passive safety systems appears to be a recurrent issue

Polarized advanced fuel reactors

International Nuclear Information System (INIS)

Kulsrud, R.M.

1987-07-01

The d- 3 He reaction has the same spin dependence as the d-t reaction. It produces no neutrons, so that if the d-d reactivity could be reduced, it would lead to a neutron-lean reactor. The current understanding of the possible suppression of the d-d reactivity by spin polarization is discussed. The question as to whether a suppression is possible is still unresolved. Other advanced fuel reactions are briefly discussed. 11 refs

Modular fabrication and characterization of complex silicon carbide composite structures Advanced Reactor Technologies (ART) Research Final Report (Feb 2015 – May 2017)

Energy Technology Data Exchange (ETDEWEB)

Khalifa, Hesham [General Atomics, San Diego, CA (United States)

2017-08-03

Advanced ceramic materials exhibit properties that enable safety and fuel cycle efficiency improvements in advanced nuclear reactors. In order to fully exploit these desirable properties, new processing techniques are required to produce the complex geometries inherent to nuclear fuel assemblies and support structures. Through this project, the state of complex SiC-SiC composite fabrication for nuclear components has advanced significantly. New methods to produce complex SiC-SiC composite structures have been demonstrated in the form factors needed for in-core structural components in advanced high temperature nuclear reactors. Advanced characterization techniques have been employed to demonstrate that these complex SiC-SiC composite structures provide the strength, toughness and hermeticity required for service in harsh reactor conditions. The complex structures produced in this project represent a significant step forward in leveraging the excellent high temperature strength, resistance to neutron induced damage, and low neutron cross section of silicon carbide in nuclear applications.

Code qualification of structural materials for AFCI advanced recycling reactors.

Energy Technology Data Exchange (ETDEWEB)

Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L. (Nuclear Engineering Division); (ORNL)

2012-05-31

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP

Research reactors; Les piles de recherche

Energy Technology Data Exchange (ETDEWEB)

Kowarski, L. [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires]|[Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)

1955-07-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

Research reactors; Les piles de recherche

Energy Technology Data Exchange (ETDEWEB)

Kowarski, L [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires; [Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)

1955-07-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

Development of Advanced High Uranium Density Fuels for Light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Blanchard, James [Univ. of Wisconsin, Madison, WI (United States); Butt, Darryl [Boise State Univ., ID (United States); Meyer, Mitchell [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Corporation, Pittsburgh, PA (United States)

2016-02-15

This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

Planning the Decommissioning of Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Podlaha, J., E-mail: pod@ujv.cz [Nuclear Research Institute Rez, 25068 Rez (Czech Republic)

2013-08-15

In the Czech Republic, three research nuclear reactors are in operation. According to the valid legislation, preliminary decommissioning plans have been prepared for all research reactors in the Czech Republic. The decommissioning plans shall be updated at least every 5 years. Decommissioning funds have been established and financial resources are regularly deposited. Current situation in planning of decommissioning of research reactors in the Czech Republic, especially planning of decommissioning of the LVR-15 research reactor is described in this paper. There appeared new circumstances having wide impact on the decommissioning planning of the LVR-15 research reactor: (1) Shipment of spent fuel to the Russian Federation for reprocessing and (2) preparation of processing of radioactive waste from reconstruction of the VVR-S research reactor (now LVR-15 research reactor). The experience from spent fuel shipment to the Russian Federation and from the process of radiological characterization and processing of radioactive waste from reconstruction of the VVR-S research reactor (now the LVR-15 research reactor) and the impact on the decommissioning planning is described in this paper. (author)

BN800: The advanced sodium cooled fast reactor plant based on close fuel cycle

International Nuclear Information System (INIS)

Wu Xingman

2011-01-01

As one of the advanced countries with actually fastest reactor technology, Russia has always taken a leading role in the forefront of the development of fast reactor technology. After successful operation of BN600 fast reactor nuclear power station with a capacity of six hundred thousand kilowatts of electric power for nearly 30 years, and after a few decades of several design optimization improved and completed on its basis, it is finally decided to build Unit 4 of Beloyarsk nuclear power station (BN800 fast reactor power station). The BN800 fast reactor nuclear power station is considered to be the project of the world's most advanced fast reactor nuclear power being put into implementation. The fast reactor technology in China has been developed for decades. With the Chinese pilot fast reactor to be put into operation soon, the Chinese model fast reactor power station has been put on the agenda. Meanwhile, the closed fuel cycle development strategy with fast reactor as key aspect has given rise to the concern of experts and decision-making level in relevant areas. Based on the experiences accumulated in many years in dealing the Sino-Russian cooperation in fast reactor technology, with reference to the latest Russian published and authoritative literatures regarding BN800 fast reactor nuclear power station, the author compiled this article into a comprehensive introduction for reference by leaders and experts dealing in the related fields of nuclear fuel cycle strategy and fast reactor technology development researches, etc. (authors)

Supervisory Control System Architecture for Advanced Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

Cetiner, Sacit M [ORNL; Cole, Daniel L [University of Pittsburgh; Fugate, David L [ORNL; Kisner, Roger A [ORNL; Melin, Alexander M [ORNL; Muhlheim, Michael David [ORNL; Rao, Nageswara S [ORNL; Wood, Richard Thomas [ORNL

2013-08-01

This technical report was generated as a product of the Supervisory Control for Multi-Modular SMR Plants project within the Instrumentation, Control and Human-Machine Interface technology area under the Advanced Small Modular Reactor (SMR) Research and Development Program of the U.S. Department of Energy. The report documents the definition of strategies, functional elements, and the structural architecture of a supervisory control system for multi-modular advanced SMR (AdvSMR) plants. This research activity advances the state-of-the art by incorporating decision making into the supervisory control system architectural layers through the introduction of a tiered-plant system approach. The report provides a brief history of hierarchical functional architectures and the current state-of-the-art, describes a reference AdvSMR to show the dependencies between systems, presents a hierarchical structure for supervisory control, indicates the importance of understanding trip setpoints, applies a new theoretic approach for comparing architectures, identifies cyber security controls that should be addressed early in system design, and describes ongoing work to develop system requirements and hardware/software configurations.

Structural and piping issues in the design certification of advanced reactors

International Nuclear Information System (INIS)

Ali, S.A.; Terao, D.; Bagchi, G.

1996-01-01

The purpose of this paper is to discuss the design certification of structures and piping for evolutionary and passive advanced light water reactors. Advanced reactor designs are based on a set of assumed site-related parameters that are selected to envelop a majority of potential nuclear power plant sites. Multiple time histories are used as the seismic design basis in order to cover the majority of potential sites in the US. Additionally, design are established to ensure that surface motions at a particular site will not exceed the enveloped standard design surface motions. State-of-the-art soil-structure interaction (SSI) analyses have been performed for the advanced reactors, which include structure-to-structure interaction for all seismic Category 1 structures. Advanced technology has been utilized to exclude the dynamic effects of pipe rupture from structural design by demonstrating that the probability of pipe rupture is extremely low. For piping design, the advanced reactor vendors have developed design acceptance criteria (DAC) which provides the piping design analysis methods, design procedures, and acceptance criteria. In SECY-93-087 the NRC staff recommended that the Commission approve the approach to eliminate the OBE from the design of structures and piping in advanced reactors and provided guidance which identifies the necessary changes to existing seismic design criteria. The supplemental criteria address fatigue, seismic anchor motion, and piping stress limits when the OBE is eliminated

Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

International Nuclear Information System (INIS)

Heeger, Karsten M.

2014-01-01

This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zero$\\theta . Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.

Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

Energy Technology Data Exchange (ETDEWEB)

Heeger, Karsten M. [Yale Univ., New Haven, CT (United States)

2014-09-13

This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zero$\\theta_{13}$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.

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

International Nuclear Information System (INIS)

Waddell, J.D.

1977-01-01

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

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

Nuclear research reactors in Brazil

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

Nuclear research reactors in Brazil

International Nuclear Information System (INIS)

Cota, Anna Paula Leite; Mesquita, Amir Zacarias

2011-01-01

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

Guidelines for Self-assessment of Research Reactor Safety

International Nuclear Information System (INIS)

2018-01-01

Self-assessment is an organization’s internal process to review its current status, processes and performance against predefined criteria and thereby to provide key elements for the organization’s continual development and improvement. Self-assessment helps the organization to think through what it is expected to do, how it is performing in relation to these expectations, and what it needs to do to improve performance, fulfil the expectations and achieve better compliance with the predefined criteria. This publication provides guidelines for a research reactor operating organization to perform a self-assessment of the safety management and the safety of the facility and to identify gaps between the current situation and the IAEA safety requirements for research reactors. These guidelines also provide a methodology for Member States, regulatory bodies and operating organizations to perform a self-assessment of their application of the provisions of the Code of Conduct on the Safety of Research Reactors. This publication also addresses planning, implementation and follow-up of actions to enhance safety and strengthen application of the Code. The guidelines are applicable to all types of research reactor and critical and subcritical assemblies, at all stages in their lifetimes, and to States, regulatory bodies and operating organizations throughout all phases of research reactor programmes. Research reactor operating organizations can use these guidelines at any time to support self-assessments conducted in accordance with the organization’s integrated management system. These guidelines also serve as a tool for an organization to prepare to receive an IAEA Integrated Safety Assessment of Research Reactors (INSARR) mission. An important result of this is the opportunity for an operating organization to identify focus areas and make safety improvements in advance of an INSARR mission, thereby increasing the effectiveness of the mission and efficiency of the

Thermal hydraulic tests for reactor safety system -Research on the improvement of nuclear safety-

International Nuclear Information System (INIS)

Chung, Moon Ki; Park, Chun Kyeong; Yang, Seon Kyu; Chung, Chang Hwan; Chun, Shee Yeong; Song, Cheol Hwa; Chun, Hyeong Gil; Chang, Seok Kyu; Chung, Heung Joon; Won, Soon Yeon; Cho, Yeong Ro; Kim, Bok Deuk; Min, Kyeong Ho

1994-07-01

The present research aims at the development of the thermal hydraulic verification test technology for the reactor safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. (Author)

Recent advances in the utilization and the irradiation technology of the refurbished BR2 reactor

International Nuclear Information System (INIS)

Dekeyser, J.; Benoit, P.; Decloedt, C.; Pouleur, Y.; Verwimp, A.; Weber, M.; Vankeerberghen, M.; Ponsard, B.

1999-01-01

Operation and utilization of the materials testing reactor BR2 at the Belgian Nuclear Research Centre (SCK·CEN) has since its start in 1963 always followed closely the needs and developments of nuclear technology. In particular, a multitude of irradiation experiments have been carried out for most types of nuclear power reactors, existing or under design. Since the early 1990s and increased focus was directed towards more specific irradiation testing needs for light water reactor fuels and materials, although other areas of utilization continued as well (e.g. fusion reactor materials, safety research, ...), including also the growing activities of radioisotope production and silicon doping. An important milestone was the decision in 1994 to implement a comprehensive refurbishment programme for the BR2 reactor and plant installations. The scope of this programme comprised very substantial studies and hardware interventions, which have been completed in early 1997 within planning and budget. Directly connected to this strategic decision for reactor refurbishment was the reinforcement of our efforts to requalify and upgrade the existing irradiation facilities and to develop advanced devices in BR2 to support emerging programs in the following fields: - LWR pressure vessel steel, - LWR irradiation assisted stress corrosion cracking (IASCC), - reliability and safety of high-burnup LWR fuel, - fusion reactor materials and blanket components, - fast neutron reactor fuels and actinide burning, - extension and diversification of radioisotope production. The paper highlights these advances in the areas of BR2 utilisation and the ongoing development activities for the required new generation of irradiations devices. (author)

Performance of the advanced cold neutron source and optics upgrades at the NIST Research Reactor

International Nuclear Information System (INIS)

Williams, R.E.; Kopetka, P.; Cook, J.C.; Rowe, J.M.

2003-01-01

On March 6, 2002, the NIST Research Reactor resumed routine operation following a six-month shutdown for facility upgrades and maintenance. During the shutdown, the original liquid hydrogen cold neutron source was removed, and the advanced cold source was installed. An optical filter was installed on one of the neutron guides, NG-3, replacing a crystal filter for the 30-m SANS instrument and the guide used between the chopper disks of the Disk Chopper time-of-flight Spectrometer (DCS) installed on NG-4 has been recently reconfigured. Additional improvements in the neutron optics of various instruments are being made. The advanced liquid hydrogen cold neutron source performs as expected, nearly doubling the flux available to most instruments. The measured gains range from about 1.4 at 2 A, to over a factor of two at 15 A. Also as expected, the heat load in the new source increased to 1200 watts, but the previously existing refrigerator has easily accommodated the increase. With intensity gains of a factor of two in the important long wavelength region of the spectrum, the advanced cold source significantly enhances the measurement capability of the cold neutron scattering instrumentation at NIST. The optical filter on NG-3 is also very successful; the 30-m SANS has an additional gain of two at 17 A. A system of refracting lenses and prisms near the SANS sample position has made possible measurements at low Q (0.0005 A -1 ) that were previously not feasible. The DCS has also seen additional intensity gain factors in excess of two for the majority of experiments and at short neutron wavelengths the gains exceed three. In addition, two new triple axis spectrometers will feature double-focusing monochromators in order to exploit the full size of the available thermal and cold neutron beam tubes. The success of the advanced cold source and enhanced neutron optics contributed to the recognition of the NIST Center for Neutron Research as 'the premiere neutron scattering

The training and research reactor of the Zittau Technical College

International Nuclear Information System (INIS)

Ackermann, G.; Hampel, R.; Konschak, K.

1979-01-01

The light-water moderated training and research reactor of the Zittau Technical College, which has been put into operation 1 July 1979, is described. Having a power of 10 MW, it is provided for education of students and advanced training of nuclear power plant staff members. High inherent nuclear safety and economy of operation are achieved by appropriate design of the reactor core and the use of fresh fuel elements provided for the 10-MW research reactor at the Rossendorf Central Institute for Nucleear Research for one year on a loan basis. Further characteristics of the reactor are easy accessibility of the core interior for in-core studies, sufficient external experimental channels, and a control and protection system meeting the requirements of teaching operation. The installed technological and dosimetric devices not only ensure reliable operation of the reactor, but also extend the potentialities of experimental work and education that is reported in detail. The principles on which the training programs are based are explained in the light of some examples. The training reactor is assumed to serve for providing basic knowledge about processes in nuclear power stations with pressurized water reactors. Where the behaviour of a nuclear power station cannot sufficiently be demonstrated by the training reactor, a reasonable completion of practical training at special simulation models and experimental facilities of the Technical College and at the nuclear power plant simulator of the Rheinsberg nuclear power plant school has been conceived. (author)

Design Requirements of an Advanced HANARO Reactor Core Cooling System

International Nuclear Information System (INIS)

Park, Yong Chul; Ryu, Jeong Soo

2007-12-01

An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

Advanced automation concepts applied to Experimental Breeder Reactor-II startup

International Nuclear Information System (INIS)

Berkan, R.C.; Upadhyaya, B.R.; Bywater, R.L.

1991-08-01

The major objective of this work is to demonstrate through simulations that advanced liquid-metal reactor plants can be operated from low power by computer control. Development of an automatic control system with this objective will help resolve specific issues and provide proof through demonstration that automatic control for plant startup is feasible. This paper presents an advanced control system design for startup of the Experimental Breeder Reactor-2 (EBR-2) located at Idaho Falls, Idaho. The design incorporates recent methods in nonlinear control with advanced diagnostics techniques such as neural networks to form an integrated architecture. The preliminary evaluations are obtained in a simulated environment by a low-order, valid nonlinear model. Within the framework of phase 1 research, the design includes an inverse dynamics controller, a fuzzy controller, and an artificial neural network controller. These three nonlinear control modules are designed to follow the EBR-2 startup trajectories in a multi-input/output regime. They are coordinated by a supervisory routine to yield a fault-tolerant, parallel operation. The control system operates in three modes: manual, semiautomatic, and fully automatic control. The simulation results of the EBR-2 startup transients proved the effectiveness of the advanced concepts. The work presented in this paper is a preliminary feasibility analysis and does not constitute a final design of an automated startup control system for EBR-2. 14 refs., 43 figs

Advanced Instrumentation and Control Methods for Small and Medium Reactors with IRIS Demonstration

Energy Technology Data Exchange (ETDEWEB)

J. Wesley Hines; Belle R. Upadhyaya; J. Michael Doster; Robert M. Edwards; Kenneth D. Lewis; Paul Turinsky; Jamie Coble

2011-05-31

Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investigate, develop, and validate advanced methods for sensing, controlling, monitoring, diagnosis, and prognosis of these reactors, and to demonstrate the methods with application to one of the proposed integral pressurized water reactors (IPWR). For this project, the IPWR design by Westinghouse, the International Reactor Secure and Innovative (IRIS), has been used to demonstrate the techniques developed under this project. The research focuses on three topical areas with the following objectives. Objective 1 - Develop and apply simulation capabilities and sensitivity/uncertainty analysis methods to address sensor deployment analysis and small grid stability issues. Objective 2 - Develop and test an autonomous and fault-tolerant control architecture and apply to the IRIS system and an experimental flow control loop, with extensions to multiple reactor modules, nuclear desalination, and optimal sensor placement strategy. Objective 3 - Develop and test an integrated monitoring, diagnosis, and prognosis system for SMRs using the IRIS as a test platform, and integrate process and equipment monitoring (PEM) and process and equipment prognostics (PEP) toolboxes. The research tasks are focused on meeting the unique needs of reactors that may be deployed to remote locations or to developing countries with limited support infrastructure. These applications will require smaller, robust reactor designs with advanced technologies for sensors, instrumentation, and control. An excellent overview of SMRs is described in an article by Ingersoll (2009). The article refers to these as deliberately small reactors. Most of these have modular characteristics, with multiple units deployed at the same plant site. Additionally, the topics focus

Research reactor`s role in Korea

Energy Technology Data Exchange (ETDEWEB)

Choi, C-O [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

1996-12-31

After a TRIGA MARK-II was constructed in 1962, new research activity of a general nature, utilizing neutrons, prevailed in Korea. Radioisotopes produced from the MARK-II played a good role in the 1960`s in educating people as to what could be achieved by a neutron source. Because the research reactor had implanted neutron science in the country, another TRIGA MARK-III had to be constructed within 10 years after importing the first reactor, due to increased neutron demand from the nuclear community. With the sudden growth of nuclear power, however, the emphasis of research changed. For a while research activities were almost all oriented to nuclear power plant technology. However, the specifics of nuclear power plant technology created a need for a more highly capable research reactor like HANARO 30MWt. HANARO will perform well with irradiation testing and other nuclear programs in the future, including: production of key radioisotopes, doping of silicon by transmutation, neutron activation analysis, neutron beam experiments, cold neutron source. 3 tabs., 2 figs.

Advanced reactor design study. Assessing nonbackfittable concepts for improving uranium utilization in light water reactors

International Nuclear Information System (INIS)

Fleischman, R.M.; Goldsmith, S.; Newman, D.F.; Trapp, T.J.; Spinrad, B.I.

1981-09-01

The objective of the Advanced Reactor Design Study (ARDS) is to identify and evaluate nonbackfittable concepts for improving uranium utilization in light water reactors (LWRs). The results of this study provide a basis for selecting and demonstrating specific nonbackfittable concepts that have good potential for implementation. Lead responsibility for managing the study was assigned to the Pacific Northwest Laboratory (PNL). Nonbackfittable concepts for improving uranium utilization in LWRs on the once-through fuel cycle were selected separately for PWRs and BWRs due to basic differences in the way specific concepts apply to those plants. Nonbackfittable concepts are those that are too costly to incorporate in existing plants, and thus, could only be economically incorporated in new reactor designs or plants in very early stages of construction. Essential results of the Advanced Reactor Design Study are summarized

Australian research reactor studies

International Nuclear Information System (INIS)

McCulloch, D.B.

1978-01-01

The Australian AEC has two research reactors at the Lucas Heights Research Establishment, a 10 HW DIDO class materials testing reactor, HIFAR, and a smaller 100kW reactor MOATA, which was recently upgraded from 10kW power level. Because of the HIFAR being some 20 years old, major renewal and repair programmes are necessary to keep it operational. To enable meeting projected increases in demand for radioisotopes, plans for a new reactor to replace the HIFAR have been made and the design criteria are described in the paper. (author)

Conceptual Nuclear Design Of Two Models Of Research Reactor Proposed For Vietnam

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Huynh Ton Nghiem; Le Vinh Vinh; Vo Doan Hai Dang

2007-01-01

The joint study on the development of a new research reactor model for Vietnam was done. The KAERI (Korea Atomic Energy Research Institute) experts and DNRI (Dalat Nuclear Research Institute) researchers developed an advanced HANARO reactor (AHR), a 20-MW open-tank-in-pool type reactor, upward cooled and moderated by light water, reflected by heavy water and rod type fuel assemblies used. Based on the AHR model, a MTR reactor with plate fuel assemblies was developed. Computer codes named MCNP and MVP/BURN were used. Major analyses have been done for the relevant nuclear design parameters such as the neutron flux and power distributions, reactivity coefficients, control rod worth, etc. in both with clean, unperturbed core and equilibrium core condition. In case of AHR model, calculation results using MVP/BURN and MCNP codes were compared with the results using HELIOS and MCNP codes by KAERI experts and they are in a good agreement. (author)

2012 review of French research reactors

International Nuclear Information System (INIS)

Estrade, Jerome

2013-01-01

Proposed by the French Reactor Operators' Club (CER), the meeting and discussion forum for operators of French research reactors, this report first gives a brief presentation of these reactors and of their scope of application, and a summary of highlights in 2012 for each of them. Then, it proposes more detailed presentations and reviews of characteristics, activities, highlights, objectives and results for the different types of reactors: neutron beam reactors (Orphee, High flux reactor-Laue-Langevin Institute or HFR-ILL), technological irradiation reactors (Osiris and Phenix), training reactors (Isis and Azur), reactors for safety research purposes (Cabri and Phebus), reactors for neutronic studies (Caliban, Prospero, Eole, Minerve and Masurca), and new research reactors (the RES facility and the Jules Horowitz reactor or JHR)

Irradiation Facilities at the Advanced Test Reactor

International Nuclear Information System (INIS)

S. Blaine Grover

2005-01-01

The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC) (formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens

Advances in Reactor Physics, Mathematics and Computation. Volume 2

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, Volume 2, are divided into 7 sessions bearing on: - session 7: Deterministic transport methods 1 (7 conferences), - session 8: Interpretation and analysis of reactor instrumentation (6 conferences), - session 9: High speed computing applied to reactor operations (5 conferences), - session 10: Diffusion theory and kinetics (7 conferences), - session 11: Fast reactor design, validation and operating experience (8 conferences), - session 12: Deterministic transport methods 2 (7 conferences), - session 13: Application of expert systems to physical aspects of reactor design and operation.

Advanced ceramic cladding for water reactor fuel

International Nuclear Information System (INIS)

Feinroth, H.

2000-01-01

Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

U.S. Status of Fast Reactor Research and Technology

International Nuclear Information System (INIS)

Hill, Robert

2012-01-01

Summary: • Fast reactor R&D is focused on key technologies innovations for performance improvement (cost reduction) and safety: 1. System Integration and Concept Development; 2. Safety Technology; 3. Advanced Materials; 4. Ultrasonic Viewing; 5. Advanced Energy Conversion (Supercritical CO 2 Brayton cycle); 6. Reactor Simulation; 7. Nuclear Data; 8. Advanced Fuels. • Fast reactors have flexible capability for actinide management: – A wide variety of fuel cycle options are being considered; • International R&D collaboration pursued in Generation-IV and multilateral arrangements

Needs for development in nondestructive testing for advanced reactor systems

International Nuclear Information System (INIS)

McClung, R.W.

1978-01-01

The needs for development of nondestructive testing (NDT) techniques and equipment were surveyed and analyzed relative to problem areas for the Liquid-Metal Fast Breeder Reactor, the Molten-Salt Breeder Reactor, and the Advanced Gas-Cooled Reactor. The paper first discusses the developmental needs that are broad-based requirements in nondestrutive testing, and the respective methods applicable, in general, to all components and reactor systems. Next, the requirements of generic materials and components that are common to all advanced reactor systems are examined. Generally, nondestructive techniques should be improved to provide better reliability and quantitativeness, improved flaw characterization, and more efficient data processing. Specific recommendations relative to such methods as ultrasonics, eddy currents, acoustic emission, radiography, etc., are made. NDT needs common to all reactors include those related to materials properties and degradation, welds, fuels, piping, steam generators, etc. The scope of applicability ranges from initial design and material development stages through process control and manufacturing inspection to in-service examination

Global trends in advanced reactor developments, and the role of the IAEA

International Nuclear Information System (INIS)

Kupitz, J.; Cleveland, J.

2000-01-01

Due to further increases in the world's population along with further industrialization and economic development, global energy demand will surely continue to increase in the 21 century. To assure that nuclear power remains a viable option in meeting energy demands in the near and medium terms, new reactor designs for all principle reactor lines and for different applications are being developed in a number of countries. Common goals for these new designs are high availability, user-friendly features, competitive economics and compliance with internationally recognized safety objectives. World-wide, considerable efforts are being made to develop advanced nuclear power Various organizations are involved, including governments, industries, utilities, universities, national laboratories, and research institutes. Expenditures for development of new designs, technology improvements, and the related research for the major reactor types combined is estimated to exceed US$ 2 billion per year. This paper gives an overview about nuclear power technology development programmes and projects in Member States and the role of the IAEA as a forum for informatics exchange and co-operative research. (author)

Advance Liquid Metal Reactor Discrete Dynamic Event Tree/Bayesian Network Analysis and Incident Management Guidelines (Risk Management for Sodium Fast Reactors)

Energy Technology Data Exchange (ETDEWEB)

Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Groth, Katrina M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-04-01

Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self-correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the system's design to manage the accident. Inherently and passively safe designs are laudable, but nonetheless extreme boundary conditions can interfere with the design attributes which facilitate inherent safety, thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayesian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The authors would like to acknowledge the U.S. Department of Energy's Office of Nuclear Energy for funding this research through Work Package SR-14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at Argonne National Laboratory, Oak Ridge National Laboratory, and Idaho National Laboratory for their continue d contributions to the advanced reactor PRA mission area.

The CEA research reactors

International Nuclear Information System (INIS)

Schwartz, J.P.

1993-01-01

Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

Advanced Small Modular Reactor Economics Model Development

Energy Technology Data Exchange (ETDEWEB)

Harrison, Thomas J [ORNL

2014-10-01

The US Department of Energy Office of Nuclear Energy’s Advanced Small Modular Reactor (SMR) research and development activities focus on four key areas: Developing assessment methods for evaluating advanced SMR technologies and characteristics; and Developing and testing of materials, fuels and fabrication techniques; and Resolving key regulatory issues identified by US Nuclear Regulatory Commission and industry; and Developing advanced instrumentation and controls and human-machine interfaces. This report focuses on development of assessment methods to evaluate advanced SMR technologies and characteristics. Specifically, this report describes the expansion and application of the economic modeling effort at Oak Ridge National Laboratory. Analysis of the current modeling methods shows that one of the primary concerns for the modeling effort is the handling of uncertainty in cost estimates. Monte Carlo–based methods are commonly used to handle uncertainty, especially when implemented by a stand-alone script within a program such as Python or MATLAB. However, a script-based model requires each potential user to have access to a compiler and an executable capable of handling the script. Making the model accessible to multiple independent analysts is best accomplished by implementing the model in a common computing tool such as Microsoft Excel. Excel is readily available and accessible to most system analysts, but it is not designed for straightforward implementation of a Monte Carlo–based method. Using a Monte Carlo algorithm requires in-spreadsheet scripting and statistical analyses or the use of add-ons such as Crystal Ball. An alternative method uses propagation of error calculations in the existing Excel-based system to estimate system cost uncertainty. This method has the advantage of using Microsoft Excel as is, but it requires the use of simplifying assumptions. These assumptions do not necessarily bring into question the analytical results. In fact, the

Studies on environment safety and application of advanced reactor for inland nuclear power plants