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Sample records for argonne advanced research reactor

  1. 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

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

    International Nuclear Information System (INIS)

    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 from normal high

  3. 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

  4. Remote dismantlement activities for the Argonne CP-5 Research Reactor

    International Nuclear Information System (INIS)

    The Department of Energy's (DOE's) Robotics Technology Development Program (RTDP) is participating in the dismantlement of a mothballed research reactor, Chicago Pile Number 5 (CP-5), at Argonne National Laboratory (ANL) to demonstrate technology developed by the program while assisting Argonne with their remote system needs. Equipment deployed for CP-5 activities includes the dual-arm work platform (DAWP), which will handle disassembly of reactor internals, and the RedZone Robotics-developed 'Rosie' remote work vehicle, which will perform size reduction of shield plugs, demolition of the biological shield, and waste packaging. Remote dismantlement tasks are scheduled to begin in February of 1997 and to continue through 1997 and beyond

  5. Specialists' meeting on advanced controls for fast reactors, Argonne, Illinois, USA June 20-22, 1989

    International Nuclear Information System (INIS)

    The Specialists' Meeting on ''Advanced Controls for Fast Reactors'' was held in Argonne, Illinois, USA, from June 20 to 22, 1989. The meeting was sponsored by the International Atomic Energy Agency on the recommendation of the IAEA International Working Group on Fast Reactors and was hosted by Argonne National Laboratory and the US Department of Energy. It was attended by 20 participants and observers from Argentina, France, Germany, Japan, India, the USSR, the United Kingdom, the United States of America, and the IAEA. The purpose of the meeting was to provide an opportunity for participating countries to review and discuss their views on design and technology for advanced control in fast reactors. During the meeting papers were presented by the participants on behalf of their countries and organizations. Presentations were followed by open discussions on the subjects covered by the papers and summaries of the discussions were drafted. After the formal sessions were completed, a final discussion session was held and summaries, general conclusions and recommendations were approved by consensus. A separate abstract was prepared for each of the 22 papers presented at this meeting. Refs, figs, tabs, diagrams and photos

  6. Advances in reactor safety research

    International Nuclear Information System (INIS)

    The Nuclear Safety Project is an important part of the German reactor safety research programme. It works on problems concerning safety and environemental risks of LWR reactors and reprocessing plants and investigates accident consequences. At the 1978 annual meeting, the core behaviour on cooling and reactivity disturbances was discussed, as well as release, retention, and possible radiological effects of radioactive pollutants. Among other subjects, fission product retention in LWR reactors and reprocessing plants were reported on as well as hypothetic core meltdown. (orig.)

  7. Advanced research reactor fuel development

    International Nuclear Information System (INIS)

    The fabrication technology of the U3Si 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 U3Si2 dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U3Si2 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 ∼ 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 γ-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 U3Si2. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano-plates will be conducted in the Advanced Test Reactor(ATR). 49

  8. 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

  9. Advanced liquid metal reactor development at Argonne National Laboratory during the 1980s

    International Nuclear Information System (INIS)

    Argonne National Laboratory's (ANL'S) effort to pursue the exploitation of liquid metal cooled reactor (LMR) characteristics has given rise to the Integral Fast Reactor (IFR) concept, and has produced substantial technical advancement in concept implementation which includes demonstration of high burnup capability of metallic fuel, demonstration of injection casting fabrication, integral demonstration of passive safety response, and technical feasibility of pyroprocessing. The first half decade of the 90's will host demonstration of the IFR closed fuel cycle technology at the prototype scale. The EBR-II reactor will be fueled with ternary alloy fuel in HT-9 cladding and ducts, and pyroprocessing and injection casting refabrication of EBR-II fuel will be conducted using near-commercial sized equipment at the Fuel cycle Facility (FCF) which is co-located adjacent to EBR-II. Demonstration will start in 1992. The demonstration of passive safety response achievable with the IFR design concept, (already done in EBR-II in 1986) will be repeated in the mid 90's using the IFR prototype recycle fuel from the FCF. The demonstration of scrubbing of the reprocessing fission product waste stream, with recycle of the transuranics to the reactor for consumption, will also occur in the mid 90's. 30 refs

  10. 'Experience with decommissioning of research and test reactors at Argonne National Laboratory'

    International Nuclear Information System (INIS)

    A large number of research reactors around the world have reached the end of their useful operational life. Many of these are kept in a controlled storage mode awaiting decontamination and decommissioning (D and D). At Argonne National Laboratory located near Chicago in the United States of America, significant experience has been gained in the D and D of research and test reactors. These experiences span the entire range of activities in D and D - from planning and characterization of the facilities to the eventual disposition of all waste. A multifaceted D nd D program has been in progress at the Argonne National Laboratory - East site for nearly a decade. The program consists of three elements: - D and D of nuclear facilities on the site that have reached the end of their useful life; - Development and demonstrations of technologies that help in safe and cost effective D and D; - Presentation of training courses in D and D practices. Nuclear reactor facilities have been constructed and operated at the ANL-E site since the earliest days of nuclear power. As a result, a number of these early reactors reached end-of-life long before reactors on other sites and were ready for D and D earlier. They presented an excellent set of test beds on which D and D practices and technologies could be demonstrated in environments that were similar to commercial reactors, but considerably less hazardous. As shown, four reactor facilities, plutonium contaminated glove boxes and hot cells, a cyclotron facility and assorted other nuclear related facilities have been decommissioned in this program. The overall cost of the program has been modest relative to the cost of comparable projects undertaken both in the U.S. and abroad. The safety record throughout the program was excellent. Complementing the actual operations, a set of D and D technologies are being developed. These include robotic methods of tool handling and operation, chemical and laser decontamination techniques, sensors

  11. Advanced fuel in the Budapest research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hargitai, T.; Vidovsky, I. [KFKI Atomic Energy Research Inst., Budapest (Hungary)

    1997-07-01

    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)

  12. China Advanced Research Reactor Project Progress in 2011

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    2011, China Advanced Research Reactor (CARR) Project finished the B stage commissioning and resolved the relative technical problems. Meanwhile, the acceptance items and the cold neutron source were carrying out.

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

    International Nuclear Information System (INIS)

    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

  14. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    International Nuclear Information System (INIS)

    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

  15. China Advanced Research Reactor Project Progress in 2012

    Institute of Scientific and Technical Information of China (English)

    ZHAO; Tie-jun

    2012-01-01

    <正>In 2012, all the commissioning for the China Advanced Research Reactor (CARR) had been finished and the diffraction pattern had been successfully obtained on the neutron scattering spectrometer. Meanwhile, the cold neutron source project and the acceptance items of CARR project had been carrying out.

  16. Environmental assessment related to the decontamination and decommissioning of the Argonne National Laboratory CP-5 research reactor

    International Nuclear Information System (INIS)

    Five alternatives for the decontamination and decommissioning of the Argonne National Labortory CP-5 research reactor are considered. Results of this study on environmental changes and impacts due to the action indicate that there will be no adverse impact on land use; decommissioning of the facility will release about 1.2 ha (3 acres) of a previously restricted area for unrestricted use, whereas radioactive-waste burial will occupy only an estimated 0.03 ha (0.07 acre. Some of the biotic habitat, vegetation, and animal life of the 1.2-ha (3-acre) waste-storage yard will be disturbed or destroyed during decontamination of the yard. The impact will be negligible in terms of the local ecosystem. There will be minimal socioeconomic impact on the area. Radiological impacts on the population from nonaccidental releases of the radionuclides 3H, 60Co, 55Fe, and 63Ni will include a dose commitment possibly as high as 0.19 mrem to the lungs of an individual working onsite and located about 100 m (300 ft) to the northeast of the reactor building. The cumulative dose to the population within an 80-km (50-mi) radius is 8.33 person-rem; this is about 10-5 of the annual natural-background dose for this area. The risks of significant radiological impacts on the population from accidents of natural catastrophies at the reactor site are extremely small. A cumulative occupational dose of about 21 person-rem will be received by the work force of up to about 50 persons participating in the dismantling activities. Population doses during the transportation of reactor scrap and wastes from dismantlement will be about 50% of the cumulative population dose within 80 km of the site. A cumulative occupational dose of about 24 person-rem could be received by the drivers of the transport trucks shipping the radioactive wastes to Richland, Washington

  17. The status of facilities at China Advanced Research Reactor

    International Nuclear Information System (INIS)

    A 60 MW research reactor, so called China Advanced Research Reactor (CARR,) was built in China Institute of Atomic Energy (CIAE), located in the southwest of Beijing and about 37 kilometers away from the central city. CARR is a tank-in-pool inverse neutron trap type reactor using D2O reflector, the designed optimal undisturbed thermal neutron flux is 8×1014 n⋅cm-2⋅s-1. A liquid D2 cold source will be equipped and the installation will be finished at the end of 2015. As a multipurpose research reactor, its main applications include neutron scattering, neutron activation analysis, isotope production, silicon doping, fuel element test, fundamental nuclear physics and so on. On March 13rd, 2012 CARR realized the 72 h stable operation with the full power. And the official operation license is expected to be issued at the beginning of next year. Cooperating with the internal and international users in the first phase ten instruments complete construction and are under commissioning, which are High Resolution Powder Diffractometer, High Intensity Powder Diffractometer, Residual Stress Diffractometer, Texture Diffractometer, Four Circle Diffractometer, Reflectometer, Small Angle Neutron Scattering, two Thermal Triple Axis Spectrometers and Isotope Separator On-Line instrument . In the second phase 7 instruments were approved and are under construction now. Although the operation license was not issued, the reactor was permitted to do the testing run several times and some results were obtained during the instrument commissioning.

  18. Chemical research at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

    Argonne National Laboratory is a research and development laboratory located 25 miles southwest of Chicago, Illinois. It has more than 200 programs in basic and applied sciences and an Industrial Technology Development Center to help move its technologies to the industrial sector. At Argonne, basic energy research is supported by applied research in diverse areas such as biology and biomedicine, energy conservation, fossil and nuclear fuels, environmental science, and parallel computer architectures. These capabilities translate into technological expertise in energy production and use, advanced materials and manufacturing processes, and waste minimization and environmental remediation, which can be shared with the industrial sector. The Laboratory`s technologies can be applied to help companies design products, substitute materials, devise innovative industrial processes, develop advanced quality control systems and instrumentation, and address environmental concerns. The latest techniques and facilities, including those involving modeling, simulation, and high-performance computing, are available to industry and academia. At Argonne, there are opportunities for industry to carry out cooperative research, license inventions, exchange technical personnel, use unique research facilities, and attend conferences and workshops. Technology transfer is one of the Laboratory`s major missions. High priority is given to strengthening U.S. technological competitiveness through research and development partnerships with industry that capitalize on Argonne`s expertise and facilities. The Laboratory is one of three DOE superconductivity technology centers, focusing on manufacturing technology for high-temperature superconducting wires, motors, bearings, and connecting leads. Argonne National Laboratory is operated by the University of Chicago for the U.S. Department of Energy.

  19. 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.

  20. Design requirement for electrical system of an advanced research reactor

    International Nuclear Information System (INIS)

    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

  1. Design approach to the development of an advanced HANARO research reactor

    International Nuclear Information System (INIS)

    Based on the experiences of the HANARO construction and operation, a project to design an advanced research reactor was launched in 2003 to prepare for the future needs of a research reactor. Many improvements identified during the HANARO operation and utilization will be incorporated into the design of the advanced research reactor. This paper deals with the basic principles of the design approach and the preliminary design features of the reactor under study

  2. The neutron texture diffractometer at the China Advanced Research Reactor

    Science.gov (United States)

    Mei-Juan, Li; Xiao-Long, Liu; Yun-Tao, Liu; Geng-Fang, Tian; Jian-Bo, Gao; Zhou-Xiang, Yu; Yu-Qing, Li; Li-Qi, Wu; Lin-Feng, Yang; Kai, Sun; Hong-Li, Wang; R. Santisteban, J.; Dong-Feng, Chen

    2016-03-01

    The first neutron texture diffractometer in China has been built at the China Advanced Research Reactor, due to strong demand for texture measurement with neutrons from the domestic user community. This neutron texture diffractometer has high neutron intensity, moderate resolution and is mainly applied to study texture in commonly used industrial materials and engineering components. In this paper, the design and characteristics of this instrument are described. The results for calibration with neutrons and quantitative texture analysis of zirconium alloy plate are presented. The comparison of texture measurements with the results obtained in HIPPO at LANSCE and Kowari at ANSTO illustrates the reliability of the texture diffractometer. Supported by National Nature Science Foundation of China (11105231, 11205248, 51327902) and International Atomic Energy Agency-TC program (CPR0012)

  3. Safety features and research needs of westinghouse advanced reactors

    International Nuclear Information System (INIS)

    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)

  4. Report of Nuclear Fusion Reactor Engineering Research Meeting. 6. Advanced reactor engineering technology for nuclear fusion demonstration reactor

    International Nuclear Information System (INIS)

    This research meeting has been held every year, and the 6th meeting was held on January 17, 1995 at University of Tokyo. As the type of a demonstration reactor, tokamak type and helical type were set up, and the topics on the various subjects of their reactor engineering technology were presented, and active discussion was carried out. At the meeting, lectures were given on the reactor engineering technology required for a prototype reactor, the material technology supposed for a demonstration reactor, thermal-electric conversion and the direct electricity generation using Nernst effect, the advanced manufacturing technology of functional, structural materials, the application of high temperature superconductors to nuclear fusion reactors, the reactor engineering technology required for a helical type demonstration reactor, and tokamak demonstration reactor and the common technology of fission and fusion. This report is the summary of these lecture materials. The useful knowledges were obtained for considering the development of nuclear fusion reactor technology hereafter in this meeting. (K.I.)

  5. Method of advancing research and development of fast breeder reactors

    International Nuclear Information System (INIS)

    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.)

  6. Advanced Reciprocating Engine Systems (ARES) Research at Argonne National Laboratory. A Report

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, Sreenath [Argonne National Lab. (ANL), Argonne, IL (United States); Biruduganti, Muni [Argonne National Lab. (ANL), Argonne, IL (United States); Bihari, Bipin [Argonne National Lab. (ANL), Argonne, IL (United States); Sekar, Raj [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-08-01

    The goals of these experiments were to determine the potential of employing spectral measurements to deduce combustion metrics such as HRR, combustion temperatures, and equivalence ratios in a natural gas-fired reciprocating engine. A laser-ignited, natural gas-fired single-cylinder research engine was operated at various equivalence ratios between 0.6 and 1.0, while varying the EGR levels between 0% and maximum to thereby ensure steady combustion. Crank angle-resolved spectral signatures were collected over 266-795 nm, encompassing chemiluminescence emissions from OH*, CH*, and predominantly by CO2* species. Further, laser-induced gas breakdown spectra were recorded under various engine operating conditions.

  7. The Advanced Photon Source: A national synchrotron radiation research facility at Argonne National Laboratory

    International Nuclear Information System (INIS)

    The vision of the APS sprang from prospective users, whose unflagging support the project has enjoyed throughout the decade it has taken to make this facility a reality. Perhaps the most extraordinary aspect of synchrotron radiation research, is the extensive and diverse scientific makeup of the user community. From this primordial soup of scientists exchanging ideas and information, come the collaborative and interdisciplinary accomplishments that no individual alone could produce. So, unlike the solitary Roentgen, scientists are engaged in a collective and dynamic enterprise with the potential to see and understand the structures of the most complex materials that nature or man can produce--and which underlie virtually all modern technologies. This booklet provides scientists and laymen alike with a sense of both the extraordinary history of x-rays and the knowledge they have produced, as well as the potential for future discovery contained in the APS--a source a million million times brighter than the Roentgen tube

  8. The Advanced Photon Source: A national synchrotron radiation research facility at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-10-01

    The vision of the APS sprang from prospective users, whose unflagging support the project has enjoyed throughout the decade it has taken to make this facility a reality. Perhaps the most extraordinary aspect of synchrotron radiation research, is the extensive and diverse scientific makeup of the user community. From this primordial soup of scientists exchanging ideas and information, come the collaborative and interdisciplinary accomplishments that no individual alone could produce. So, unlike the solitary Roentgen, scientists are engaged in a collective and dynamic enterprise with the potential to see and understand the structures of the most complex materials that nature or man can produce--and which underlie virtually all modern technologies. This booklet provides scientists and laymen alike with a sense of both the extraordinary history of x-rays and the knowledge they have produced, as well as the potential for future discovery contained in the APS--a source a million million times brighter than the Roentgen tube.

  9. Argonne to open new facility for advanced vehicle testing

    CERN Multimedia

    2002-01-01

    Argonne National Laboratory will open it's Advanced Powertrain Research Facility on Friday, Nov. 15. The facility is North America's only public testing facility for engines, fuel cells, electric drives and energy storage. State-of-the-art performance and emissions measurement equipment is available to support model development and technology validation (1 page).

  10. Advancing nuclear technology and research. The advanced test reactor national scientific user facility

    International Nuclear Information System (INIS)

    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 INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation 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. The mission of the ATR NSUF is to provide access to world-class facilities, thereby facilitating the advancement of nuclear science and technology. Cost free access to the ATR, INL post irradiation examination facilities, and partner 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 United States Department of Energy. To increase overall research capability, ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. (author)

  11. Advanced Reactor Safety Research Program quarterly report, April--June 1977. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-11-01

    Information is presented concerning accident energetics; core debris behavior; sodium containment and structural integrity; research for elevated temperature design criteria; fuel motion detection; ACPR fuel motion system; and advanced reactor safety research assessment.

  12. Conjugate heat transfer simulations of advanced research reactor fuel

    International Nuclear Information System (INIS)

    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

  13. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    Energy Technology Data Exchange (ETDEWEB)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials

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

    International Nuclear Information System (INIS)

    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)

  15. Technical Research on Safety Management and Effective Application of China Advanced Research Reactor

    International Nuclear Information System (INIS)

    China Advanced Research Reactor (CARR) is a tank in pool type, light water cooled, heavy water reflected research reactor. The maximum thermal neutron flux of the reactor is 1.0x1015 cm-2s-1, and the reactor power is 60 MW. The reactor was designed and constructed completely by China Institute of Atomic Energy (CIAE). The construction project began on Aug. 26, 2002, reactor criticality was achieved on May 13, 2010, and it is scheduled to complete power increasing tests by the end of 2011. Future operation of CARR is preparing and its utilization program is considered. It is expected that CARR will greatly improve and enhance the comprehensive research capability of nuclear science and technology and push the peaceful use of nuclear technology forward. The paper briefly presents the reactor safety features, the operation organization and responsibilities, the management of operation safety, and the future utilizations. According to national safety regulations of research reactor, evaluation of operation safety of CARR shall be executed after initial operation at power level and submit the revised ''Final Safety Analysis Report'' (FSAR) to the regulatory body.Ordinary operation shall be approved and operation license shall be issued by the regulatory body after review on the ''Final Safety Analysis Report.'' Vertical and horizontal channels with associated equipment and instruments are installed in reactor core and in heavy water reflector. CARR will be used to produce variety of RIs in comprehensive fields, to meet the requirements of engineering tests and irradiation for developing NPP fuels and materials in China, to apply for NTD of mono-crystalline silicone, NAA, neutron photography and to provide high intense neutron beam for application of neutron scattering experiments in an adequate scale and others, etc. (author)

  16. Russian RERTR program: Advanced LEU fuel development for research reactors

    International Nuclear Information System (INIS)

    Pin-type fuel elements of new generation have been developed for conversion of Russian-built research reactors to LEU. This paper contains the main results of out-of-pile investigations during manufacturing of experimental batches of mini- and full size fuel elements for irradiation tests in 'MIR' (Dmitrovgrad) and VVR-M (Gatchina) reactor respectively. U-Mo LEU fuel granules have been manufactured both by centrifugal atomization and by hydride-dehydride process. Loading of mini-fuel elements made 4 and 6 g U/cm3, whereas full size ones were loaded to 5.3 g U/cm3. All fuel elements went through quality control of cladding thickness and uniformity of fuel granules distribution through the length of each one. (author)

  17. Advanced sodium fast reactor accident source terms : research needs.

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard [IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France; Denning, Richard [Ohio State University, Columbus, OH; Ohno, Shuji [Japan Atomic Energy Agency, Ibaraki, Japan; Zeyen, Roland [Institute for Energy Petten, Saint-Paul-lez-Durance, France

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  18. A Preliminary Calculation of Annular Core Design for a High-flux Advanced Research Reactor

    International Nuclear Information System (INIS)

    Many of research reactors in operation over the world become old and the number of research reactors is expected to be reduced around 1/3 within a next decade. So it may be necessary to prepare in advance for the future demands of research reactors with a high performance. Therefore, based on the HANARO experiences through design to operation, a concept development of an improved research reactor is under doing. In this paper, 10 MW conceptual annular core is proposed and its basic characteristics were analyzed as a preliminary step

  19. Conceptual design of second safety system for an advanced research reactor

    International Nuclear Information System (INIS)

    In recent IAEA recommends that at least one automatic shutdown system shall be incorporated in the design of research reactor, and a second independent shutdown system shall be considered and may be required, depending on the characteristics of the reactor. It would be required to reflect the strengthening of IAEA's safety policy to developing project of an advanced research reactor which was based on the experiences in HANARO construction and operation. The reflector dump system for an advanced research reactor is required for safety system diversity. The safety system shuts the reactor down to a safe sub-critical state by dropping hafnium control absorber rods into the reactor core and by dumping the inner annular volume of the D2O reflector tank into the dump tank

  20. Conceptual design of second safety system for an advanced research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Sang Ik; Park, Y. C.; Cho, Y. G.; Lee, J. H.; Ryu, J. S.; Park, C

    2005-01-01

    In recent IAEA recommends that at least one automatic shutdown system shall be incorporated in the design of research reactor, and a second independent shutdown system shall be considered and may be required, depending on the characteristics of the reactor. It would be required to reflect the strengthening of IAEA's safety policy to developing project of an advanced research reactor which was based on the experiences in HANARO construction and operation. The reflector dump system for an advanced research reactor is required for safety system diversity. The safety system shuts the reactor down to a safe sub-critical state by dropping hafnium control absorber rods into the reactor core and by dumping the inner annular volume of the D{sub 2}O reflector tank into the dump tank.

  1. Research reactors - an overview

    Energy Technology Data Exchange (ETDEWEB)

    West, C.D.

    1997-03-01

    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.

  2. Simulation research and optimal design for digital power regulating system of China advanced research reactor

    International Nuclear Information System (INIS)

    Based on SimPort simulation platform of nuclear power plant, a simulation model for Digital Power Regulating System (DPRS) of China Advanced Research Reactor (CARR) was established. By simulating the transient state of DPRS using this model, the adjusting parameters for the digital PID controller were determined. According to the features of the driving mechanism, the effects of the driving accuracy of the control rod and the displacement delay between electromagnetic coil and armature upon system stability and the regulating performance were analyzed, furthermore, their stability limit values were obtained respectively. The research results of this paper have some engineering practical value. (authors)

  3. Method of advancing research and development of fast breeder reactors

    International Nuclear Information System (INIS)

    In the long term plan of atomic energy development and utilization, FBRs are to be developed as the main of future nuclear power generation in Japan, and when the development is advanced, it is positivity aimed at building up the plutonium utilization system using FBRs superior to the uranium utilization system with LWRs. Also it was decided that it is necessary to exert incessant effort for the development of FBRs under the proper cooperation system of the government and people for a considerable long period, and as for the concrete development, hereafter, the deliberation is advanced by the expert subcommittee on FBR development project of the Atomic Energy Commission in succession. The subcommittee was founded in May, 1986, to carry out the deliberation on the long term promotion measures for the development of FBRs, the promotion measures for the research and development, the evaluation and examination of the basic specification of a demonstration FBR, the promotion measures for the international cooperation and other important matters related to the development of FBRs. The construction of the prototype FBR 'Monju' is in progress aiming at the criticality in 1992, and the start of construction of a demonstration FBR is expected in the latter half of 1990s. The situation around the development of FBRs, the fundamentals for promoting the research and development, and the subjects of the research and development are reported. (Kako, I.)

  4. Design guide on instrumentation and control system for an advanced research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Ki; Jung, H. S.; Choi, Y. S.; Ryu, J. S.; Park, C

    2004-12-01

    Instrumentation and control system is to regulate reactor power automatically and to perform protective functions to shut down the reactor safely in case of an accident threatening the reactor safety. If the protective action would not be properly activated, an abnormal situation might be connected to a significant radiation hazard, even though a research reactor is maintained at low pressure and low temperature condition. For this reason, the reactor protection system for the advanced research reactor will be designed as per the safety design requirements for nuclear power generating stations. System classification are suggested based on the HANARO concept. Considering an international growing tendency toward nuclear safety enhancement and supplement of the existing one, a revised classification system will be applied to the components, structures and systems for the advanced research reactor. The reactor protection system is classified to the safety class and the others are non-safety grade. Based on the HANARO experience, the design guide aims at the improvement of reliability of safety and achievement of automatic control and signal processing by digitalization. To satisfy functional diversity requirement, the first shutdown system adopts a concept of control rod with a stepping motor while the second shutdown system uses D{sub 2}O dump system. The first shutdown system will use a PLC-type digital logic processor but the second shutdown system will use a conventional relay logic for the concept of equipment diversity. The reactor protective function for the advanced research reactor has a higher reliability by using two different shutdown mechanism. Although most of the instrumentation systems, except the reactor protection system are classified to non-safety grade, some systems like reactor regulation system or engineered safety feature actuation system selectively adopts the safety design requirements in order to ensure the safety-related function. All of the

  5. Simulation Research on Neutron Guide System CNGC for China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; HE; Lin-feng; WANG; Yu; WANG; Hong-li; LIU; Yun-tao; CHEN; Dong-feng; ZHAO; Zhi-xiang

    2012-01-01

    <正>The out-pile section of the neutron guide CNGC at CARR (China Advanced Research Reactor) was designed by Monte Carlo simulation with VITESS. The out-pile section of CNGC will be spitted to CNGC-S and CNGC-N, the cold neutron imaging facility and small angle neutron scattering facility will be installed at the end of guides respectively. XRD patterns of Bi1-xLaxFe1-yScyO3 were shown in Fig. 1.

  6. Reactor D and D at Argonne National Laboratory - lessons learned

    International Nuclear Information System (INIS)

    This paper focuses on the lessons learned during the decontamination and decommissioning (D and D) of two reactors at Argonne National Laboratory-East (ANL-E). The Experimental Boiling Water Reactor (EBWR) was a 100 MW(t), 5 MSV(e) proof-of-concept facility. The Janus Reactor was a 200 kW(t) reactor located at the Biological Irradiation Facility and was used to study the effects of neutron radiation on animals

  7. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  8. Research reactors

    International Nuclear Information System (INIS)

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world's research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted

  9. Simulation research and optimal design for digital power regulating system of China advanced research reactor

    International Nuclear Information System (INIS)

    Based on SimPort simulation platform of nuclear power plant, a simulation model for Digital Power Regulating System (DPRS) of China Advanced Research Reactor (CARR) was established. The transient state of DPRS was simulation studied using this model. According to the characteristics of the driving mechanism of the control rods, the effects of the driving precision of the control rod and its displacement delay upon the system stability were analyzed. Considering the process requirements of CARR and the function characteristic of DRPS, the adjusting parameters for the digital PID controller and the stability limits of the driving mechanism of the control rods were obtained. The sampling period of the digital PID controller is 100 ms and its proportion gain is 300. The stability limit of the driving precision of the control rod is 0.4 mm. The stability limit of displacement delay between electromagnetic coil and armature is 6.0 mm. (authors)

  10. Research reactors

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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 21st 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 D2O reflector not less than 8 x 1014 n/cm2. 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 1015 n/cm2. 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.)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  14. Status of the Advanced Photon Source at Argonne National Laboratory

    International Nuclear Information System (INIS)

    The Advanced Photon Source at Argonne National Laboratory is a third-generation light source optimized for production of high-brilliance undulator radiation in the hard x-ray portion of the spectrum. A user community representing all major centers of synchrotron research, including universities, industry, and federal laboratories, will utilize these x-ray beams for investigations across a diverse range of disciplines. All technical facilities and components required for operations have been completed and installed, and are well along in the commissioning process. Major design goals and Department of Energy milestones have been met or exceeded. Project funds have been maximized to construct a number of beamline components and user facilities over and above those called for in the original project scope. Research teams preparing experimental apparatus at the Advanced Photon Source have procured strong funding support. copyright 1996 American Institute of Physics

  15. Preliminary Design of a Reflector Dump System for an Advanced HANARO Research Reactor

    International Nuclear Information System (INIS)

    Recently, the IAEA has recommended that at least one automatic shutdown system shall be incorporated in the design of a research reactor, and a second independent shutdown system shall be considered and may be required, depending on the characteristics of the reactor. According to the recommend, it would be required to reflect the strengthening of the IAEA's safety policy for developing a project of an AHRR (Advanced HANARO Research Reactor) which is based on the experiences in HANARO construction and operation. The reflector dump system for an AHRR is required for safety system diversity. It is considered that the safety system of an AHRR shuts the reactor down to a safe sub-critical state by dropping hafnium control absorber rods into the reactor core and by dumping the volume of the D2O reflector tank into the dump tank. This paper summarizes the preliminary design on the heavy water dump system which may be considered as a second safety system for a new research reactor

  16. Experimental software design of neutron texture diffractometer at China advanced research reactor

    International Nuclear Information System (INIS)

    The experimental software of the neutron texture diffractometer at China Advanced Research Reactor (CARR) was designed. Based on the principle of texture measurement by neutron diffraction and the motion control and data acquisition system of the diffractometer, the functions needed for texture measurement were proposed. Then the flow charts of these functions were described in detail and realized by Python language in Linux system. The experimental software for CARR neutron texture diffractometer has been successfully accomplished. (authors)

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

    International Nuclear Information System (INIS)

    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 21st century. (author)

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

    International Nuclear Information System (INIS)

    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)

  19. Current status of advanced pelletized cold moderators development for IBR-2M research reactor

    International Nuclear Information System (INIS)

    The world's first advanced pelletized cold neutron moderator is prepared to be put into operation at the IBR-2M pulsed research reactor. It provides long-wavelength neutrons to the most of neutron spectrometers at the beams of the IBR-2M reactor. Aromatic hydrocarbons are used as a material for cold moderators. It is a very attractive material because of its high radiation resistance, good moderating properties, incombustibility, etc. It is shown that the idea of beads transport by a helium flow at cryogenic temperatures is successful. The recent progress and plans for moderator development at the IBR-2M reactor as well as the experimental results of beads transport are discussed in the paper

  20. Current status of advanced pelletized cold moderators development for IBR-2M research reactor

    Science.gov (United States)

    Kulikov, S.; Belyakov, A.; Bulavin, M.; Mukhin, K.; Shabalin, E.; Verhoglyadov, A.

    2013-03-01

    The world's first advanced pelletized cold neutron moderator is prepared to be put into operation at the IBR-2M pulsed research reactor. It provides long-wavelength neutrons to the most of neutron spectrometers at the beams of the IBR-2M reactor. Aromatic hydrocarbons are used as a material for cold moderators. It is a very attractive material because of its high radiation resistance, good moderating properties, incombustibility, etc. It is shown that the idea of beads transport by a helium flow at cryogenic temperatures is successful. The recent progress and plans for moderator development at the IBR-2M reactor as well as the experimental results of beads transport are discussed in the paper.

  1. Argonne National Laboratory: An example of a US nuclear research centre

    International Nuclear Information System (INIS)

    The nuclear era was ushered in 1942 with the demonstration of a sustained nuclear chain reaction in Chicago Pile 1 facility. The USA then set up five large national multi disciplinary laboratories for developing nuclear technology for civilian use and three national laboratories for military applications. Reactor development, including prototype construction, was the main focus of the Argonne National Laboratory. More than 100 power reactors operating in the USA have benefited from R and D in the national laboratories. However, currently the support for nuclear power has waned. With the end of the cold war there has also been a need to change the mission of laboratories involved in military applications. For all laboratories of the Department of Energy (DOE) the mission, which was clearly focused earlier on high risk, high payoff long term R and D has now become quite diffused with a number of near term programmes. Cost and mission considerations have resulted in shutting down of many large facilities as well as auxiliary facilities. Erosion of infrastructure has also resulted in reduced opportunities for research which means dwindling of interest in nuclear science and engineering among the younger generation. The current focus of nuclear R and D in the DOE laboratories is on plant life extension, deactivation and decommissioning, spent fuel management and waste management. Advanced aspects include space nuclear applications and nuclear fusion R and D. At the Argonne National Laboratory, major initiatives for the future would be in the areas of science, energy, environment and non-proliferation technologies. International collaboration would be useful mechanisms to achieve cost effective solutions for major developmental areas. These include reactor operation and safety, repositories for high level nuclear waste, reactor system decommissioning, large projects like a nuclear fusion reactor and advanced power reactors. The IAEA could have a positive role in these

  2. US Research Programme to Support Advanced Reactors and Fuel Cycle Options

    International Nuclear Information System (INIS)

    Nuclear energy is an essential element of the United States of America’s clean energy portfolio. The US Department of Energy’s Office of Nuclear Energy is addressing the challenges to the expansion of civilian nuclear energy through a science based research and development (R&D) programme. This paper describes our research in the areas of advanced reactor technology and fuel cycle options, and our engagements with key stakeholders including international partners, private industry, the regulator and the university community. (author)

  3. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    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

  4. 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.

  5. Three-dimensional flow field analysis of the standard fuel assembly for China advanced research reactor

    International Nuclear Information System (INIS)

    Numerical simulation of the flow field of the standard fuel assembly in China Advanced Research Reactor is carried out by using computational fluid dynamics software CFX4.4 and CFX5.5. The flow distribution and pressure difference of different coolant channels in the standard fuel assembly at rated operating condition are reached. Based on the computational pressure drop results of different flow rate, the resistance characteristic curve is given and compared with experimental results. The two results are in good agreement. (authors)

  6. Construction and adjustment of neutron texture diffractometer at China advanced research reactor

    International Nuclear Information System (INIS)

    The neutron texture diffractometer is one of important and commonly used neutron instruments in the international neutron scattering laboratories. Under the demands of texture measurement with neutrons from domestic user community, the neutron texture diffractometer has been built at China Advanced Research Reactor (CARR). Currently, the preliminary adjustment and calibration with neutrons for this instrument has been finished. In this paper, the measurement principle and advantages of neutron texture diffractometer were briefly introduced. The key components and detailed characteristics for neutron texture diffractometer at CARR and the corresponding results of calibration and performance test were also presented. (authors)

  7. 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".

  8. Design of the cold neutron triple-axis spectrometer at the China Advanced Research Reactor

    Science.gov (United States)

    Cheng, P.; Zhang, Hongxia; Bao, W.; Schneidewind, A.; Link, P.; Grünwald, A. T. D.; Georgii, R.; Hao, L. J.; Liu, Y. T.

    2016-06-01

    The design of the first cold neutron triple-axis spectrometer at the China Advanced Research Reactor is presented. Based on the Monte Carlo simulations using neutron ray-tracing program McStas, the parameters of major neutron optics in this instrument are optimized. The neutron flux at sample position is estimated to be 5.6 ×107 n/cm2/s at neutron incident energy Ei=5 meV when the reactor operates normally at the designed 60 MW power. The performances of several neutron supermirror polarizing devices are compared and their critical parameters are optimized for this spectrometer. The polarization analysis will be realized with a flexible switch from the unpolarized experimental mode.

  9. Verification Survey of the Building 315 Zero Power Reactor-6 Facility, Argonne National Laboratory-East, Argonne, Illinois

    International Nuclear Information System (INIS)

    Oak Ridge Institute for Science and Education (ORISE) conducted independent verification radiological survey activities at Argonne National Laboratory's Building 315, Zero Power Reactor-6 facility in Argonne, Illinois. Independent verification survey activities included document and data reviews, alpha plus beta and gamma surface scans, alpha and beta surface activity measurements, and instrumentation comparisons. An interim letter report and a draft report, documenting the verification survey findings, were submitted to the DOE on November 8, 2006 and February 22, 2007, respectively (ORISE 2006b and 2007). Argonne National Laboratory-East (ANL-E) is owned by the U.S. Department of Energy (DOE) and is operated under a contract with the University of Chicago. Fundamental and applied research in the physical, biomedical, and environmental sciences are conducted at ANL-E and the laboratory serves as a major center of energy research and development. Building 315, which was completed in 1962, contained two cells, Cells 5 and 4, for holding Zero Power Reactor (ZPR)-6 and ZPR-9, respectively. These reactors were built to increase the knowledge and understanding of fast reactor technology. ZPR-6 was also referred to as the Fast Critical Facility and focused on fast reactor studies for civilian power production. ZPR-9 was used for nuclear rocket and fast reactor studies. In 1967, the reactors were converted for plutonium use. The reactors operated from the mid-1960's until 1982 when they were both shut down. Low levels of radioactivity were expected to be present due to the operating power levels of the ZPR's being restricted to well below 1,000 watts. To evaluate the presence of radiological contamination, DOE characterized the ZPRs in 2001. Currently, the Melt Attack and Coolability Experiments (MACE) and Melt Coolability and Concrete Interaction (MCCI) Experiments are being conducted in Cell 4 where the ZPR-9 is located (ANL 2002 and 2006). ANL has performed final

  10. Speech by Ichiro Miyanaga, Executive Director, Japan Atomic Energy Research Institute at the 1984 international meeting on Reduced Enrichment for Research and Test Reactors, Argonne National Laboratory, October 17, 1984

    International Nuclear Information System (INIS)

    Full text: It is a great pleasure for me to extend my greetings and best wishes to all of you on this honorable occasion. I would like to take this opportunity to express my appreciation to Argonne National Laboratory for their continued support and cooperation through Joint Study with Japan Atomic Energy Research Institute as well as Kyoto University. Japan Atomic Energy Research Institute has been endeavoring to convert the present HEU fuels of research reactors to MEU fuels, as addressed by Prof. Kanda of Kyoto University in the first session of this meeting. For TRR-2 and JMTR, MEU cores are expected to attain their first criticality early the year 1986. Recognizing the final goal of RERTR program lies in using LEU fuels, we will soon start examinations and tests in JMTR for LEU fuel development according to the current feasibility study. The full core demonstration of JMTR with LEU silicide fuel would be expected in 1990. At the same time, Japan Atomic Energy Research Institute is engaged in-JRR-3M Construction Project. JRR-3M was originally designed to use MEU fuels. However, in consideration of recent progress in LEU fuel-technology, the design was changed to using LEU fuels. JRR-3M is scheduled to begin operation in 1989, which will be one of the first high performance research reactors with LEU fuels in the world. For implementing MEU and LEU Program, we have the so-called Five Agency Committee, the members of which are Science and Technology Agency, Ministry of Foreign Affairs, Ministry of Education, Science and Culture, and the direct implementors, Kyoto University and JAERI. The government members in this committee have played an important role for promoting the Program by review and consultation. Most concern we have now is about the stable supply of MEU for a necessary period and the reprocessing of LEU fuels which are the final goal in our reactors. For this reason, I would like to ask the U.S. Government to meet these requirements for us to perform

  11. Advanced Fission Reactor Program objectives

    International Nuclear Information System (INIS)

    The objective of an advanced fission reactor program should be to develop an economically attractive, safe, proliferation-resistant fission reactor. To achieve this objective, an aggressive and broad-based research and development program is needed. Preliminary work at Brookhaven National Laboratory shows that a reasonable goal for a research program would be a reactor combining as many as possible of the following features: (1) initial loading of uranium enriched to less than 15% uranium 235, (2) no handling of fuel for the full 30-year nominal core life, (3) inherent safety ensured by core physics, and (4) utilization of natural uranium at least 5 times as efficiently as light water reactors

  12. DOE/NE University Program in robotics for advanced reactors research

    International Nuclear Information System (INIS)

    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)

  13. Advanced reactor licensing issues

    International Nuclear Information System (INIS)

    In July 1986 the US Nuclear Regulatory Commission issued a Policy Statement on the Regulation of Advanced Nuclear Power Plants. As part of this policy advanced reactor designers were encouraged to interact with NRC 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). As a result of these interactions certain safety issues associated with these advanced reactor designs have been identified as key to the licensability of the designs as proposed by DOE. The major issues in this regard are: (1) selection and treatment of accident scenarios; (2) selection of siting source term; (3) performance and reliability of reactor shutdown and decay heat removal systems; (4) need for conventional containment; (5) need for conventional emergency evacuation; (6) role of the operator; (7) treatment of balance of plant; and (8) modular approach. This paper provides a status of the NRC review effort, describes the above issues in more detail and provides the current status and approach to the development of licensing guidance on each

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

    International Nuclear Information System (INIS)

    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

  15. 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

  16. Design and manufacture of neutron time of flight spectrometer on China Advanced Research Reactor

    International Nuclear Information System (INIS)

    The cold or thermal neutron energy spectra on China Advanced Research Reactor (CARR) could be directly measured by neutron time of flight spectrometer. Spectrometer structure and selected parameters of its key components were introduced. The impact of chopper slit and flux limit slit on neutron counts and pulse width was analyzed. The formulas of neutron counts and pulse width which were dependent on neutron wavelength were acquired. According to neutron energy spectrum measurement requirement for high fluence rate neutron beam, low-sensitivity detector, detector flux limit slit and multi-channel scaler for data acquisition were selected. These would ensure that the count loss rate was less than 0.5%. Electronics framework of detection system was designed and the total resolution time was 22.15-29.46 μs. (authors)

  17. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  20. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, K.C. [Univ. of Turabo, Gurabo, Puerto (Puerto Rico). College of Engineering; Yahr, G.T. [Oak Ridge National Lab., TN (United States)

    1995-08-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in K{sub Q} due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail.

  1. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    International Nuclear Information System (INIS)

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in KQ due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail

  2. 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)

    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

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

    International Nuclear Information System (INIS)

    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

  4. Advances and highlights of the CNEA qualification program as high density fuel manufacturer for research reactors

    International Nuclear Information System (INIS)

    One of the main objectives of CNEA regarding the fuel for research reactors is the development and qualification of the manufacturing of LEU high-density fuels. The qualification programs for both types of fuels, Silicide fuel and U- x Mo fuel, are similar. They include the following activities: development and set up of the fissile compound manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of mini plates and plates, design and fabrication of fuel assembly prototypes for irradiation, post-irradiation examination and feedback for manufacturing improvements. This paper describes the different activities performed within each program during the last year and the main advances and achievements of the programs within this period. The main achievements may be summarized in the following activities: Continuation of the irradiation of the first silicide fuel element in the R A3. Completion of the manufacturing of the second silicide fuel element, licensing and beginning of its irradiation in the R A3. Development of the HMD Process to manufacture U-Mo powder (pUMA project). Set up of fuel plates manufacturing at industrial level using U-Mo powder. Preliminary studies and the design for the irradiation of mini plates, plates and full scale fuel elements with U-Mo and 7 g U/cm3. PIE destructive studies for the P-04 silicide fuel prototype (accurate burnup determination through chemical analysis, metallography and SEM of samples from the irradiated fuel plates). Improvement and development of new characterization techniques for high density fuel plates quality control including US testing and densitometric analysis of X-ray examinations. The results obtained in this period are encouraging and also allow to foresee a wider participation of CNEA in the international effort to qualify U-Mo as a new material for the manufacturing of research reactor fuels. (author)

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

    International Nuclear Information System (INIS)

    • 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

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

    International Nuclear Information System (INIS)

    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

  7. Argonne National Laboratory research offers clues to Alzheimer's plaques

    CERN Multimedia

    2003-01-01

    Researchers from Argonne National Laboratory and the University of Chicago have developed methods to directly observe the structure and growth of microscopic filaments that form the characteristic plaques found in the brains of those with Alzheimer's Disease (1 page).

  8. Verification Survey of the Building 315 Zero Power Reactor-6 Facility, Argonne National Laboratory-East, Argonne, Illinois

    Energy Technology Data Exchange (ETDEWEB)

    W. C. Adams

    2007-05-25

    Oak Ridge Institute for Science and Education (ORISE) conducted independent verification radiological survey activities at Argonne National Laboratory’s Building 315, Zero Power Reactor-6 facility in Argonne, Illinois. Independent verification survey activities included document and data reviews, alpha plus beta and gamma surface scans, alpha and beta surface activity measurements, and instrumentation comparisons. An interim letter report and a draft report, documenting the verification survey findings, were submitted to the DOE on November 8, 2006 and February 22, 2007, respectively (ORISE 2006b and 2007).

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

    International Nuclear Information System (INIS)

    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

  10. Safety, technical and economic objectives of the electric power research institute's advanced light water reactor programme

    International Nuclear Information System (INIS)

    In the early 1980s, the Electric Power Research Institute entered into a programme to develop the safety, technical, operational and economic objectives of the next generation of LWRs. The programme benefited from the support and participation of US and international utilities, the United States Department of Energy, US reactor vendors and architect-engineering firms. This programme provided a foundation for the Nuclear Power Oversight Committee's comprehensive initiative for revitalizing nuclear power in the USA, as set forth in its 'Strategic Plan for Building New Nuclear Power Plants', published in November 1990 and updated annually each November. This Strategic Plan contains 14 'building blocks', each of which is considered essential to constructing a new nuclear plants. Building from the perspective that the Advanced Light Water Reactor (ALWR) should reflect lessons learned from the design, construction and operation of existing nuclear power plants, utilities were surveyed to help identify the cornerstone principles of what was subsequently called the 'ALWR Utility Requirements Document' (URD). The paper provides a review of the objectives underlying these principles, and the design targets that derive from this. Some examples are provided of how these targets are being met in the design of the four ALWR plants currently undergoing United States Nuclear Regulatory Commission licensing certification under United States Code of Federal Regulations 10CFR52. Some areas requiring further investigation are also described. The paper concludes by discussing key challenges to the URD targets and process, in particular with respect to regulatory and economic considerations, and provides brief highlights of steps being taken to respond to these challenges. (author). 2 tabs

  11. Project plan for the decontamination and decommissioning of the Argonne National Laboratory Experimental Boiling Water Reactor

    International Nuclear Information System (INIS)

    In 1956, the Experimental Boiling Water Reactor (EBWR) Facility was first operated at Argonne National Laboratory (ANL) as a test reactor to demonstrate the feasibility of operating an integrated power plant using a direct cycle boiling water reactor as a heat source. In 1967, ANL permanently shut down the EBWR and placed it in dry lay-up. This project plan presents the schedule and organization for the decontamination and decommissioning of the EBWR Facility which will allow it to be reused by other ANL scientific research programs. The project total estimated cost is $14.3M and is projected to generate 22,000 cubic feet of low-level radioactive waste which will be disposed of at an approved DOE burial ground. 18 figs., 3 tabs

  12. Thirty meters small angle neutron scattering instrument at China advanced research reactor

    International Nuclear Information System (INIS)

    A high resolution 30 m small angle neutron scattering (SANS) instrument has been constructed by the Institute of Chemistry, Chinese Academy of Sciences (ICCAS), and installed at China Advanced Research Reactor (CARR). It is equipped with a mechanical velocity selector, pinhole (including multi-pinhole) collimation system, sample chamber, and high resolution two dimensional 3He position sensitive neutron detector. The flexible variations of incident neutron wavelength, source to sample distance, sample to detector distance and the presence of neutron focusing lenses enable a wide Q range from 0.001 Å−1 to 0.5 Å−1 in reciprocal space and to optimize the resolution required. The instrument is the first SANS instrument in China, and can be widely used for the structure characterization of various materials, as well as kinetic and dynamic observation during external stimulation. The design and characteristics of the instrument are presented in the manuscript. -- Highlights: • The first small angle neutron scattering instrument in China is developed. • It is equipped with a pinhole, multi-pinhole and focusing lenses. • It provides flexible variations of source-sample and sample-detector distances. • A wide Q ranges from 0.001 Å−1 to 0.5 Å−1 in reciprocal space

  13. Nuclear research reactors

    International Nuclear Information System (INIS)

    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.)

  14. Thermal hydraulic R and D of Chinese advanced reactors

    International Nuclear Information System (INIS)

    The Chinese government sponsors a program of research, development, and demonstration related to advanced reactors, both small modular reactors and larger systems. These advanced reactors encompass innovative reactor concepts, such as CAP1400 - Chinese large advanced passive pressurized water reactor, Hualong one - Chinese large advanced active and passive pressurized water reactor, ACP100 - Chinese small modular reactor, SCWR- R and D of super critical water-cooled reactor in China, CLEAR - Chinese lead-cooled fast reactor, TMSR - Chinese Thorium molten-salt reactor. The thermal hydraulic R and D of those reactors are summarised. (J.P.N.)

  15. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    International Nuclear Information System (INIS)

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors

  16. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [Brookhaven National Lab., Upton, NY (United States)] [comp.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

  17. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  18. Status report on the Advanced Photon Source Project at Argonne National Laboratory

    International Nuclear Information System (INIS)

    The Advanced Photon Source at Argonne National Laboratory is designed as a national synchrotron radiation user facility which will provide extremely bright, highly energetic x-rays for multidisciplinary research. When operational, the Advanced Photon Source will accelerate positrons to a nominal energy of 7 GeV. The positrons will be manipulated by insertion devices to produce x-rays 10,000 times brighter than any currently available for research. Accelerator components, insertion devices, optical elements, and optical-element cooling schemes have been and continue to be the subjects of intensive research and development. A call for Letters of Intent from prospective users of the Advanced Photon Source has resulted in a substantial response from industrial, university, and national laboratory researchers

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

    International Nuclear Information System (INIS)

    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

  20. Research Nuclear Reactors

    International Nuclear Information System (INIS)

    Published in English and in French, this large report first proposes an overview of the use and history of research nuclear reactors. It discusses their definition, and presents the various types of research reactors which can be either related to nuclear power (critical mock-ups, material test reactors, safety test reactors, training reactors, prototypes), or to research (basic research, industry, health), or to specific particle physics phenomena (neutron diffraction, isotope production, neutron activation, neutron radiography, semiconductor doping). It reports the history of the French research reactors by distinguishing the first atomic pile (ZOE), and the activities and achievements during the fifties, the sixties and the seventies. It also addresses the development of instrumentation for research reactors (neutron, thermal, mechanical and fission gas release measurements). The other parts of the report concern the validation of neutronics calculations for different reactors (the EOLE water critical mock-up, the MASURCA air critical mock-up dedicated to fast neutron reactor study, the MINERVE water critical mock-up, the CALIBAN pulsed research reactor), the testing of materials under irradiation (OSIRIS reactor, laboratories associated with research reactors, the Jules Horowitz reactor and its experimental programs and related devices, irradiation of materials with ion beams), the investigation of accident situations (on the CABRI, Phebus, Silene and Jules Horowitz reactors). The last part proposes a worldwide overview of research reactors

  1. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    International Nuclear Information System (INIS)

    Conversion of research and test reactor cores from the use of high enrichment uranium to the use of low enrichment uranium depends on the cooperation of many research organizations, reactor operators, and government agencies. At a technical level, it involves almost all aspects of the fuel cycle, including fuel development, testing, shipping and reprocessing; experiment performance; economics; and safety and licensing aspects. The reactors involved and the conversion activities are distributed among approximately 25 countries, making this a subject which is best dealt with on an international basis. To foster direct communication in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the fifth of a series which began in 1978. The previous meetings were held at Argonne (International Meeting of Research Reactor Fuel Designers, Developers, and Fabricators, Argonne National Laboratory, Argonne, U.S.A., November 910, 1978), at Saclay (IAEA Consultants' Meeting on Research Reactor Core Conversions from HEU to LEU, Centre d'etudes Nucleaires de Saclay, Saclay, France, December 12-14, 1979), at Argonne (International Meeting on Development, Fabrication and Application of Reduced Enrichment Fuels for Research and Test Reactors, Argonne National Laboratory, Argonne, U.S.A., November 12-14, 1980) and at Juelich (Seminar on Research Reactor Operation and Use, Juelich Nuclear Research Center, Juelich, F.R.G., September 48, 1981). Proceedings from the two most recent previous meetings were published as ANL/RERTR/TM-3 (CONF-801144) and IAEA-SR-77. The spirit of this meeting differs slightly from that of the previous meetings. The advances which have been made and the growing maturity of the effort have caused a gradual shift of emphasis away from those topics which dominated the floor during the first meetings, such as fuel and methods development, and towards topics which concern more

  2. Advanced Test Reactor National Scientific User Facility Partnerships

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-03-01

    Wisconsin-Madison; (8) Illinois Institute of Technology (IIT) Materials Research Collaborative Access Team (MRCAT) beamline at Argonne National Laboratory's Advanced Photon Source; and (9) Nanoindenter in the University of California at Berkeley (UCB) Nuclear Engineering laboratory Materials have been analyzed for ATR NSUF users at the Advanced Photon Source at the MRCAT beam, the NIST Center for Neutron Research in Gaithersburg, MD, the Los Alamos Neutron Science Center, and the SHaRE user facility at Oak Ridge National Laboratory (ORNL). Additionally, ORNL has been accepted as a partner facility to enable ATR NSUF users to access the facilities at the High Flux Isotope Reactor and related facilities.

  3. Advanced Test Reactor National Scientific User Facility Partnerships

    International Nuclear Information System (INIS)

    -Madison; (8) Illinois Institute of Technology (IIT) Materials Research Collaborative Access Team (MRCAT) beamline at Argonne National Laboratory's Advanced Photon Source; and (9) Nanoindenter in the University of California at Berkeley (UCB) Nuclear Engineering laboratory Materials have been analyzed for ATR NSUF users at the Advanced Photon Source at the MRCAT beam, the NIST Center for Neutron Research in Gaithersburg, MD, the Los Alamos Neutron Science Center, and the SHaRE user facility at Oak Ridge National Laboratory (ORNL). Additionally, ORNL has been accepted as a partner facility to enable ATR NSUF users to access the facilities at the High Flux Isotope Reactor and related facilities.

  4. Technical support for the EPRI [Electric Power Research Institute] debris coolability requirement for advanced light water reactors

    International Nuclear Information System (INIS)

    The Electric Power Research Institute (EPRI) advanced light water reactor (ALWR) design requirement for cavity sizing to promote long-term core debris coolability specifies that the reactor containment be arranged to afford a floor area for potential core debris spreading of at least 0.02 square meters per thermal megawatt of reactor power. This requirement has substantial experimental and analytical support as a sufficient basis to ensure long-term debris coolability. A logical framework and presentation of such support is provided in this document. The evidence as presented demonstrates that, following vessel failure, core debris in the cavity would consist of a mixture of fragments and a relatively continuous, but porous and cracked, phase that would be distributed relatively uniformly over the area that would be available per the requirement. Considering pertinent operative heat transfer phenomena, it is also demonstrated that the heat removal rate from core debris in this configuration would be sufficient to ensure debris coolability in the reactor cavity, even when the cavity is flooded after vessel failure. 29 refs., 6 figs., 8 tabs

  5. Survey of research reactors

    International Nuclear Information System (INIS)

    A survey of reasearch reactors based on the IAEA Nuclear Research Reactor Data Base (RRDB) was done. This database includes information on 273 operating research reactors ranging in power from zero to several hundred MW. From these 273 operating research reactors 205 reactors have a power level below 5 MW, the remaining 68 reactors range from 5 MW up to several 100 MW thermal power. The major reactor types with common design are: Siemens Unterrichtsreaktors, 1.2 Argonaut reactors, Slowpoke reactors, the miniature neutron source reactors, TRIGA reactors, material testing reactors and high flux reactors. Technical data such as: power, fuel material, fuel type, enrichment, maximum neutron flux density and experimental facilities for each reactor type as well as a description of their utilization in physics and chemistry, medicine and biology, academic research and teaching, training purposes (students and physicists, operating personnel), industrial application (neutron radiography, silicon neutron transmutation doping facilities) are provided. The geographically distribution of these reactors is also shown. As conclusions the author discussed the advantages (low capital cost, low operating cost, low burn up, simple to operate, safe, less restrictive containment and sitting requirements, versatility) and disadvantages (lower sensitivity for NAA, limited radioisotope production, limited use of neutron beams, limited access to the core, licensing) of low power research reactors. 24 figs., refs. 15, Tab. 1 (nevyjel)

  6. Advanced control research at the Pennsylvania State University

    International Nuclear Information System (INIS)

    This paper summarizes several current control research projects at the Pennsylvania State University which have potential application for improving nuclear power plant operation. This research is grouped into 3 broad categories: (1) intelligent control including automated decision making, coordinating control, and reconfigurable control, (2) advanced algorithms using optimal, robust, fuzzy, and neural network control, and (3) distributed implementation with commercially available microprocessor-based controllers. These research activities encompass theoretical development coupled with experimental verification. Much of the early experimental verification and demonstration has been conducted using simulation experiments. However, current work is now beginning to include physical verification using the Penn State TRIGA nuclear research reactor and the Experimental Breeder Reactor (EBR-11) power plant operated by the Argonne National Laboratory at the Idaho National Engineering Laboratory

  7. Safeguarding research reactors

    International Nuclear Information System (INIS)

    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

  8. Decontamination and decommissioning of the JANUS reactor at the Argonne National Laboratory-East site

    International Nuclear Information System (INIS)

    Argonne National Laboratory has begun the decontamination and decommissioning (D ampersand D) of the JANUS Reactor Facility. The project is managed by the Technology Development Division's D ampersand D Program personnel. D ampersand D procedures are performed by sub-contractor personnel. Specific activities involving the removal, size reduction, and packaging of radioactive components and facilities are discussed

  9. Research nuclear reactors

    International Nuclear Information System (INIS)

    Since the divergence of the first nuclear reactor in 1942, about 600 research or test reactors have been built throughout the world. Today 255 research reactors are operating in 57 countries and about 70% are over 25 years old. Whereas there are very few reactor types for power plants because of rationalization and standardisation, there is a great diversity of research reactors. We can divide them into 2 groups: heavy water cooled reactors and light water moderated reactors. Heavy water cooled reactors are dedicated to the production of high flux of thermal neutrons which are extracted from the core by means of neutronic channels. Light water moderated reactors involved pool reactors and slightly pressurized closed reactors, they are polyvalent but their main purposes are material testing, technological irradiations, radionuclide production and neutron radiography. At the moment 8 research reactors are being built in Canada, Germany, Iran, Japan, Kazakhstan, Morocco, Russia and Slovakia and 8 others are planned in 7 countries (France, Indonesia, Nigeria, Russia, Slovakia, Thailand and Tunisia. Different research reactors are described: Phebus, Masurca, Phenix and Petten HFR. The general principles of nuclear safety applied to test reactors are presented. (A.C.)

  10. Argonne National Laboratory annual report of Laboratory Directed Research and Development Program Activities FY 2009.

    Energy Technology Data Exchange (ETDEWEB)

    Office of the Director

    2010-04-09

    I am pleased to submit Argonne National Laboratory's Annual Report on its Laboratory Directed Research and Development (LDRD) activities for fiscal year 2009. Fiscal year 2009 saw a heightened focus by DOE and the nation on the need to develop new sources of energy. Argonne scientists are investigating many different sources of energy, including nuclear, solar, and biofuels, as well as ways to store, use, and transmit energy more safely, cleanly, and efficiently. DOE selected Argonne as the site for two new Energy Frontier Research Centers (EFRCs) - the Institute for Atom-Efficient Chemical Transformations and the Center for Electrical Energy Storage - and funded two other EFRCs to which Argonne is a major partner. The award of at least two of the EFRCs can be directly linked to early LDRD-funded efforts. LDRD has historically seeded important programs and facilities at the lab. Two of these facilities, the Advanced Photon Source and the Center for Nanoscale Materials, are now vital contributors to today's LDRD Program. New and enhanced capabilities, many of which relied on LDRD in their early stages, now help the laboratory pursue its evolving strategic goals. LDRD has, since its inception, been an invaluable resource for positioning the Laboratory to anticipate, and thus be prepared to contribute to, the future science and technology needs of DOE and the nation. During times of change, LDRD becomes all the more vital for facilitating the necessary adjustments while maintaining and enhancing the capabilities of our staff and facilities. Although I am new to the role of Laboratory Director, my immediate prior service as Deputy Laboratory Director for Programs afforded me continuous involvement in the LDRD program and its management. Therefore, I can attest that Argonne's program adhered closely to the requirements of DOE Order 413.2b and associated guidelines governing LDRD. Our LDRD program management continually strives to be more efficient. In

  11. Argonne National Laboratory annual report of Laboratory Directed Research and Development Program Activities FY 2009

    International Nuclear Information System (INIS)

    I am pleased to submit Argonne National Laboratory's Annual Report on its Laboratory Directed Research and Development (LDRD) activities for fiscal year 2009. Fiscal year 2009 saw a heightened focus by DOE and the nation on the need to develop new sources of energy. Argonne scientists are investigating many different sources of energy, including nuclear, solar, and biofuels, as well as ways to store, use, and transmit energy more safely, cleanly, and efficiently. DOE selected Argonne as the site for two new Energy Frontier Research Centers (EFRCs) - the Institute for Atom-Efficient Chemical Transformations and the Center for Electrical Energy Storage - and funded two other EFRCs to which Argonne is a major partner. The award of at least two of the EFRCs can be directly linked to early LDRD-funded efforts. LDRD has historically seeded important programs and facilities at the lab. Two of these facilities, the Advanced Photon Source and the Center for Nanoscale Materials, are now vital contributors to today's LDRD Program. New and enhanced capabilities, many of which relied on LDRD in their early stages, now help the laboratory pursue its evolving strategic goals. LDRD has, since its inception, been an invaluable resource for positioning the Laboratory to anticipate, and thus be prepared to contribute to, the future science and technology needs of DOE and the nation. During times of change, LDRD becomes all the more vital for facilitating the necessary adjustments while maintaining and enhancing the capabilities of our staff and facilities. Although I am new to the role of Laboratory Director, my immediate prior service as Deputy Laboratory Director for Programs afforded me continuous involvement in the LDRD program and its management. Therefore, I can attest that Argonne's program adhered closely to the requirements of DOE Order 413.2b and associated guidelines governing LDRD. Our LDRD program management continually strives to be more efficient. In addition to

  12. Advanced fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tomita, Yukihiro [National Inst. for Fusion Science, Toki, Gifu (Japan)

    2003-04-01

    The main subjects on fusion research are now on D-T fueled fusion, mainly due to its high fusion reaction rate. However, many issues are still remained on the wall loading by the 14 MeV neutrons. In the case of D-D fueled fusion, the neutron wall loading is still remained, though the technology related to tritium breeding is not needed. The p-{sup 6}Li and p-{sup 11}B fueled fusions are not estimated to be the next generation candidate until the innovated plasma confinement technologies come in useful to achieve the high performance plasma parameters. The fusion reactor of D-{sup 3}He fuels has merits on the smaller neutron wall loading and tritium handling. However, there are difficulties on achieving the high temperature plasma more than 100 keV. Furthermore the high beta plasma is needed to decrease synchrotron radiation loss. In addition, the efficiency of the direct energy conversion from protons coming out from fusion reaction is one of the key parameters in keeping overall power balance. Therefore, open magnetic filed lines should surround the plasma column. In this paper, we outlined the design of the commercial base reactor (ARTEMIS) of 1 GW electric output power configured by D-{sup 3}He fueled FRC (Field Reversed Configuration). The ARTEMIS needs 64 kg of {sup 3}He per a year. On the other hand, 1 million tons of {sup 3}He is estimated to be in the moon. The {sup 3}He of about 10{sup 23} kg are to exist in gaseous planets such as Jupiter and Saturn. (Y. Tanaka)

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

    International Nuclear Information System (INIS)

    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

  14. 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

  15. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The overall objective of the Phase 1 effort was to demonstrate the technical feasibility of the Advanced Carbothermal Electric (ACE) Reactor concept. Unlike...

  16. Advanced Carbothermal Electric Reactor Project

    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...

  17. Advanced reactor safety research quarterly report, October-December 1982. Volume 24

    Energy Technology Data Exchange (ETDEWEB)

    None

    1984-04-01

    This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  18. 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)

    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

  19. Multipurpose research reactors

    International Nuclear Information System (INIS)

    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

  20. Advanced reactor safety research. Quarterly report, April-June 1982. Volume 22

    Energy Technology Data Exchange (ETDEWEB)

    None

    1983-10-01

    Overall objective of this work is to provide NRC a comprehensive data base essential to (1) defining key safety issues, (2) understanding risk-significant accident sequences, (3) developing and verifying models used in safety assessments, and (4) assuring the public that power reactor systems will not be licensed and placed in commercial service in the United States without appropriate consideration being given to their effects on health and safety. This report describes progress in a number of activities dealing with current safety issues relevant to both light water and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents, and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  1. 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)

    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)

  2. 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)

    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)

  3. Applications of Research Reactors

    International Nuclear Information System (INIS)

    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

  4. Research reactors in Argentina

    International Nuclear Information System (INIS)

    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

  5. Ageing of research reactors

    International Nuclear Information System (INIS)

    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)

  6. Reactor Materials Research

    International Nuclear Information System (INIS)

    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)

  7. 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)

  8. Argonne National Laboratory Annual Report of Laboratory Directed Research and Development Program Activities for FY 1994

    Energy Technology Data Exchange (ETDEWEB)

    None

    1995-02-25

    The purposes of Argonne's Laboratory Directed Research and Development (LDRD) Program are to encourage the development of novel concepts, enhance the Laboratory's R and D capabilities, and further the development of its strategic initiatives. Projects are selected from proposals for creative and innovative R and D studies which are not yet eligible for timely support through normal programmatic channels. Among the aims of the projects supported by the Program are establishment of engineering proof-of-principle; assessment of design feasibility for prospective facilities; development of an instrumental prototype, method, or system; or discovery in fundamental science. Several of these projects are closely associated with major strategic thrusts of the Laboratory as described in Argonne's Five-Year Institutional Plan, although the scientific implications of the achieved results extend well beyond Laboratory plans and objectives. The projects supported by the Program are distributed across the major programmatic areas at Argonne as indicated in the Laboratory's LDRD Plan for FY 1994. Project summaries of research in the following areas are included: (1) Advanced Accelerator and Detector Technology; (2) X-ray Techniques for Research in Biological and Physical Science; (3) Nuclear Technology; (4) Materials Science and Technology; (5) Computational Science and Technology; (6) Biological Sciences; (7) Environmental Sciences: (8) Environmental Control and Waste Management Technology; and (9) Novel Concepts in Other Areas.

  9. Vibration considerations in the design of the Advanced Photon Source at Argonne National Laboratory

    International Nuclear Information System (INIS)

    The Advanced Photon Source (APS), a new synchrotron radiation facility being built at Argonne National Laboratory, will provide the world's most brilliant X-ray beams for research in a wide range of technical fields. Successful operation of the APS requires an extremely stable positron closed orbit. Vibration of the storage ring quadrupole magnets, even in the submicron range, can lead to distortion of the positron closed orbit and to potentially unacceptable beam emittance growth, which results in degraded performance. This paper presents an overview of the technical approach used to minimize vibration response, beginning at the conceptual stage, through design and construction, and on to successful operation. Acceptance criteria relating to maximum allowable quadrupole magnet vibration are discussed. Soil properties are used to determine resonant frequencies of foundations and to predict attenuation characteristics. Two sources are considered to have the potential to excite the foundation: far-field sources, which are produced external to the facility, and near-field sources, which are produced within the facility. Measurements of ambient ground motion, monitored to determine far- field excitation, are presented. Ambient vibration was measured at several operating facilities within Argonne to gain insight on typical near-field excitation sources. Discussion covers the dynamic response characteristics of a prototypic magnet support structure to various excitations, including ambient floor motion, coolant flow, and magnet power. 19 refs., 10 figs., 5 tabs

  10. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  11. Directions in advanced reactor technology

    International Nuclear Information System (INIS)

    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

  12. Fast reactor research in Switzerland

    International Nuclear Information System (INIS)

    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)

  13. INVAP's Research Reactor Designs

    International Nuclear Information System (INIS)

    INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors IAEA safety

  14. ICONE-4: Proceedings. Volume 2: Advanced reactors

    International Nuclear Information System (INIS)

    The proceedings for this conference are contained in 5 volumes. This volume is divided into the following areas: advanced reactor requirements; advanced reactor design and analysis; arrangement and construction; specific reactor designs; demonstration testing; safety systems and analysis; component demonstration testing; advanced reactor containment design; licensing topics and updates; accelerator applications and spallation sources; and advanced reactor development. Separate abstracts were prepared for most papers in this volume

  15. Argonne National Laboratory Annual Report of Laboratory Directed Research and Development program activities FY 2011.

    Energy Technology Data Exchange (ETDEWEB)

    (Office of The Director)

    2012-04-25

    As a national laboratory Argonne concentrates on scientific and technological challenges that can only be addressed through a sustained, interdisciplinary focus at a national scale. Argonne's eight major initiatives, as enumerated in its strategic plan, are Hard X-ray Sciences, Leadership Computing, Materials and Molecular Design and Discovery, Energy Storage, Alternative Energy and Efficiency, Nuclear Energy, Biological and Environmental Systems, and National Security. The purposes of Argonne's Laboratory Directed Research and Development (LDRD) Program are to encourage the development of novel technical concepts, enhance the Laboratory's research and development (R and D) capabilities, and pursue its strategic goals. projects are selected from proposals for creative and innovative R and D studies that require advance exploration before they are considered to be sufficiently developed to obtain support through normal programmatic channels. Among the aims of the projects supported by the LDRD Program are the following: establishment of engineering proof of principle, assessment of design feasibility for prospective facilities, development of instrumentation or computational methods or systems, and discoveries in fundamental science and exploratory development.

  16. Argonne National Laboratory Annual Report of Laboratory Directed Research and Development program activities FY 2010.

    Energy Technology Data Exchange (ETDEWEB)

    (Office of The Director)

    2012-04-25

    As a national laboratory Argonne concentrates on scientific and technological challenges that can only be addressed through a sustained, interdisciplinary focus at a national scale. Argonne's eight major initiatives, as enumerated in its strategic plan, are Hard X-ray Sciences, Leadership Computing, Materials and Molecular Design and Discovery, Energy Storage, Alternative Energy and Efficiency, Nuclear Energy, Biological and Environmental Systems, and National Security. The purposes of Argonne's Laboratory Directed Research and Development (LDRD) Program are to encourage the development of novel technical concepts, enhance the Laboratory's research and development (R and D) capabilities, and pursue its strategic goals. projects are selected from proposals for creative and innovative R and D studies that require advance exploration before they are considered to be sufficiently developed to obtain support through normal programmatic channels. Among the aims of the projects supported by the LDRD Program are the following: establishment of engineering proof of principle, assessment of design feasibility for prospective facilities, development of instrumentation or computational methods or systems, and discoveries in fundamental science and exploratory development.

  17. The replacement research reactor

    International Nuclear Information System (INIS)

    As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described

  18. Advanced Reactor Development in the United States

    International Nuclear Information System (INIS)

    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

  19. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  20. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris oe National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately

  1. Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor

    International Nuclear Information System (INIS)

    The aging of research reactors worldwide has resulted in a heightened awareness in the international technical decommissioning community of the timeliness to review and address the needs of these research institutes in planning for and eventually performing the decommissioning of these facilities. By using the reactors already undergoing decommissioning as test beds for evaluating enhanced or new/innovative technologies for decommissioning, it is possible that new techniques could be made available for those future research reactor decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the research institutes in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research reactor decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to use in the research reactors. The decommissioning of the CP-5 Research Reactor is currently in the final phase of dismantlement. In this paper the authors present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors

  2. Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharyya, S. K.; Boing, L. E.

    2000-02-17

    The aging of research reactors worldwide has resulted in a heightened awareness in the international technical decommissioning community of the timeliness to review and address the needs of these research institutes in planning for and eventually performing the decommissioning of these facilities. By using the reactors already undergoing decommissioning as test beds for evaluating enhanced or new/innovative technologies for decommissioning, it is possible that new techniques could be made available for those future research reactor decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the research institutes in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research reactor decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to use in the research reactors. The decommissioning of the CP-5 Research Reactor is currently in the final phase of dismantlement. In this paper the authors present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors.

  3. Overview on New Research Reactors in China

    International Nuclear Information System (INIS)

    In China, 2 research reactors are now under construction. Correspondingly, this paper consists of 2 parts. Part 1 will focus on China Advanced Research Reactor (CARR), the reactor characteristics, utilization, safety related systems and other main systems will be described in this part. Part 2 will focus on China Experiment Fast Reactor(CEFR), the general design and the safety features in particular will be illustrated in this part. (author)

  4. Materials for advanced water cooled reactors

    International Nuclear Information System (INIS)

    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

  5. Safety of research reactors

    International Nuclear Information System (INIS)

    The number of research reactors that have been constructed worldwide for civilian applications is about 651. Of the reactors constructed, 284 are currently in operation, 258 are shut down and 109 have been decommissioned. More than half of all operating research reactors worldwide are over thirty years old. During this long period of time national priorities have changed. Facility ageing, if not properly managed, has a natural degrading effect. Many research reactors face concerns with the obsolescence of equipment, lack of experimental programmes, lack of funding for operation and maintenance and loss of expertise through ageing and retirement of the staff. Other reactors of the same vintage maintain effective ageing management programmes, conduct active research programmes, develop and retain high calibre personnel and make important contributions to society. Many countries that operate research reactors neither operate nor plan to operate power reactors. In most of these countries there is a tendency not to create a formal regulatory body. A safety committee, not always independent of the operating organization, may be responsible for regulatory oversight. Even in countries with nuclear power plants, a regulatory regime differing from the one used for the power plants may exist. Concern is therefore focused on one tail of a continuous spectrum of operational performance. The IAEA has been sending missions to review the safety of research reactors in Member States since 1972. Some of the reviews have been conducted pursuant to the IAEA' functions and responsibilities regarding research reactors that are operated within the framework of Project and Supply Agreements between Member States and the IAEA. Other reviews have been conducted upon request. All these reviews are conducted following procedures for Integrated Safety Assessment of Research Reactors (INSARR) missions. The prime objective of these missions has been to conduct a comprehensive operational safety

  6. Research reactor support

    International Nuclear Information System (INIS)

    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

  7. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  8. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    International Nuclear Information System (INIS)

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs

  9. Fuel for advanced CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor system has proven itself to be a world leader in terms of station availability and low total unit energy cost. In 1985 for example, four of the top ten reactor units in the world were CANDU reactors operating in South Korea and Canada. This excellent operating record requires an equivalent performance record of the low-cost, natural uranium fuel. Future CANDU reactors will be an evolution of the present design. Engineering work is under way to refine the existing CANDU 600 and to incorporate state-of-the-art technology, reducing the capital cost and construction schedule. In addition, a smaller CANDU 300 plant has been designed using proven CANDU 600 technology and components but with an innovative new plant layout that makes it cost competitive with coal fired plants. For the long term, work on advanced fuel cycles and major system improvements is underway ensuring that CANDU plants will stay competitive well into the next century

  10. Evaluation of research reactors

    International Nuclear Information System (INIS)

    The present status of research reactors with highly enriched (93%) uranium fuel at JAERI, JRR-2 and JMTR is described. JRR-2 is a heterogeneous type of reactor, using heavy water as moderator and coolant. It uses both MTR type and cylindrical type of fuel elements. The maximum thermal power and the thermal neutron flux are 10 MW and 2x1014 n/cm2 see respectively. The reactor has been used for various experiments such as solid state physics, material irradiation, reactor fuel irradiation and radioisotope production. The JMTR is a multi-purpose tank type material testing reactor, and light water moderator and coolant, operated at 50 MW. The evaluation of lower enriched fuel and its consequences for both reactors is considered more especially

  11. Status of the back-end optional advanced research reactor fuel development in Korea

    International Nuclear Information System (INIS)

    U-Mo fuel development has been carried out for a reactor upgrade of HANARO and the back-end option in Korea. The 2nd irradiation test of the U-Mo dispersion rod fuels is underway in HANARO in order to find the optimum uranium loading density and to investigate the applicability of the monolithic U-Mo ring fuel as well as other parameters such as particle size and cladding surface-treatment. The optical observation using an immersion camera showed that the cladding surfaces of the two U3Si and U-Mo fuels with a high power rate changed in to the darker color, which is not as severe as those of the driving fuels in HANARO. Presumably it would be acceptable. The other fuels were observed as maintaining their initial good conditions. In connection with monolithic U-Mo fuel development, some achievements such as preliminary U-Mo tube production by a continuous casting process and a successful U-Mo foil production using a roll casting process have been obtained. In addition, some investigation on the surface-treatment of multilayer coating and Zr sputtering coating has showed the possibility of eliminating the problem of a temperature rise due to the corrosion layer formation having quite a low conductivity. The next irradiation test will aim mainly at the qualification of the U-Mo dispersion fuel for HANARO around the end of next year. In the 3rd irradiation fuel bundle, some fuels related to the basic investigation tests for the monolithic U-Mo fuel and surface-treatment for anticorrosion will be loaded. (author)

  12. Thai research reactor

    International Nuclear Information System (INIS)

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  13. Advances in heavy water reactors

    International Nuclear Information System (INIS)

    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

  14. Cermet-fueled reactors for advanced space applications

    International Nuclear Information System (INIS)

    Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel were carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper

  15. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    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 1014 n/cm2/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

  16. TRIGA research reactors

    International Nuclear Information System (INIS)

    TRIGA (Training, Research, Isotope production, General-Atomic) has become the most used research reactor in the world with 65 units operating in 24 countries. The original patent for TRIGA reactors was registered in 1958. The success of this reactor is due to its inherent level of safety that results from a prompt negative temperature coefficient. Most of the neutron moderation occurs in the nuclear fuel (UZrH) because of the presence of hydrogen atoms, so in case of an increase of fuel temperature, the neutron spectrum becomes harder and neutrons are less likely to fission uranium nuclei and as a consequence the power released decreases. This inherent level of safety has made this reactor fit for training tool in university laboratories. Some recent versions of TRIGA reactors have been designed for medicine and industrial isotope production, for neutron therapy of cancers and for providing a neutron source. (A.C.)

  17. Dossier: research reactors

    International Nuclear Information System (INIS)

    Research reactors are used at the CEA (the French atomic energy commission) since many years. Their number has been reduced but they remain unique tools that CEA valorize continuously. The results of the programs involving such reactors are of prime importance for the operation of Electricite de France (EdF) park of existing power plants but also for the design of future nuclear power plants and future research reactors. This dossier presents three examples of research reactors in use at the CEA: Osiris and Orphee (CEA-Saclay), devoted to nuclear energy and fundamental research, respectively, and the critical mockups Eole, Minerve and Masurca (CEA-Cadarache) devoted to nuclear data libraries and neutronic calculation. (J.S.)

  18. Frontiers: Research highlights 1946-1996 [50th Anniversary Edition. Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    This special edition of 'Frontiers' commemorates Argonne National Laboratory's 50th anniversary of service to science and society. America's first national laboratory, Argonne has been in the forefront of U.S. scientific and technological research from its beginning. Past accomplishments, current research, and future plans are highlighted.

  19. 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.

  20. Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    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.

  1. Decontamination and dismantlement of the JANUS Reactor at Argonne National Laboratory-East. Project final report

    International Nuclear Information System (INIS)

    The decontamination and dismantlement of the JANUS Reactor at Argonne National Laboratory-East (ANL-E) was completed in October 1997. Descriptions and evaluations of the activities performed and analyses of the results obtained during the JANUS D and D Project are provided in this Final Report. The following information is included: objective of the JANUS D and D Project; history of the JANUS Reactor facility; description of the ANL-E site and the JANUS Reactor facility; overview of the D and D activities performed; description of the project planning and engineering; description of the D and D operations; summary of the final status of the JANUS Reactor facility based upon the final survey results; description of the health and safety aspects of the project, including personnel exposure and OSHA reporting; summary of the waste minimization techniques utilized and total waste generated by the project; and summary of the final cost and schedule for the JANUS D and D Project

  2. Advances in reactor physics education: Visualization of reactor parameters

    International Nuclear Information System (INIS)

    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)

  3. GE's advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    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

  4. Research Reactor Benchmarks

    International Nuclear Information System (INIS)

    A criticality benchmark experiment performed at the Jozef Stefan Institute TRIGA Mark II research reactor is described. This experiment and its evaluation are given as examples of benchmark experiments at research reactors. For this reason the differences and possible problems compared to other benchmark experiments are particularly emphasized. General guidelines for performing criticality benchmarks in research reactors are given. The criticality benchmark experiment was performed in a normal operating reactor core using commercially available fresh 20% enriched fuel elements containing 12 wt% uranium in uranium-zirconium hydride fuel material. Experimental conditions to minimize experimental errors and to enhance computer modeling accuracy are described. Uncertainties in multiplication factor due to fuel composition and geometry data are analyzed by sensitivity analysis. The simplifications in the benchmark model compared to the actual geometry are evaluated. Sample benchmark calculations with the MCNP and KENO Monte Carlo codes are given

  5. Development of the nuclear fuel materials for research reactor -Development of the advanced nuclear materials-

    International Nuclear Information System (INIS)

    This project was carried out in order to localize the fuel production and to develop the atomized uranium silicide fuels. In the fabrication part, the fuel meat was successfully cladded with good contact between meat and clad through extrusion die modification and properly controlling of extrusion parameters. The electron beam welding parameters were established properly so that the strength of welding zone showed to be same as the clad strength. The fabricating and assembling technologies of the fuel assembly have been developed and applied to the fabrication of dummy fuels. The development of coating technology on the surface of graphite crucible reduced the carbon contamination in atomized powder. In fuel performance testing part, the thermal reaction swelling tests were carried out. The results showed that the atomized fuel meat swells 30 - 40 % lower than pulverized fuel meat. The corrosion test revealed that the thickness of oxide layer ranges within the criteria of fuel safe requirement. In the fuel design part, the design criteria have been established with evaluating KMRR fuel material characteristics with respect to KMRR operation. In the hydraulic testing part, the modification of test facility and the development of the measuring instrument and data aquisition system were performed. The preliminary testing results with dummy fuels fabricated by our research team showed that the developed system works normally and the results is almost same as AECL results. (Author)

  6. Materials Engineering Research Facility (MERF)

    Data.gov (United States)

    Federal Laboratory Consortium — Argonne?s Materials Engineering Research Facility (MERF) enables engineers to develop manufacturing processes for producing advanced battery materials in sufficient...

  7. First Algerian research reactor

    International Nuclear Information System (INIS)

    In 1985, both the Algerian Commissariat of New Energies and the Argentine National Atomic Energy Commission plus the firm INVAP S.E., started a series of mutual visits aimed at defining the mechanisms for cooperation in the nuclear field. Within this framework, a commercial contract was undersigned covering the supply of a low-power reactor (RUN), designed for basic and applied research in the fields of reactor physics and nuclear engineering. The reactor may also be used for performing experiences with neutron beams, for the irradiation of several materials and for the training of technicians, scientists and operators

  8. Development Program of the Advanced HANARO Reactor in Korea

    International Nuclear Information System (INIS)

    The development program of an advanced HANARO (AHR) reactor started in Korea to keep abreast of the increasing future demand, from both home and abroad, for research activities. This paper provides a review of the status of research reactors in Korea, the operating experience of the HANARO, the design principles and preliminary features of an advanced HANARO reactor, and the specific strategy of an advanced HANARO reactor development program. The design principles were established in order to design a new multi-purpose research reactor that is safe, economically competitive and technically feasible. These include the adaptation of the HANARO design concept, its operating experience, a high ratio of flux to power, a high degree of safety, improved economic efficiency, improved operability and maintainability, increased space and expandability, and ALARA design optimization. The strategy of an advanced HANARO reactor development program considers items such as providing a digital advanced HANARO reactor in cyber space, a method for the improving the design quality and economy of research reactors by using Computer Integrated Engineering, and more effective advertising using diverse virtual reality. This development program will be useful for promoting the understanding of and interest in the operating HANARO as well as an advanced HANARO reactor under development in Korea. It will provide very useful information to a country that may need a research reactor in the near future for the promotion of public health, bio-technology, drug design, pharmacology, material processing, and the development of new materials. (author)

  9. The replacement research reactor

    International Nuclear Information System (INIS)

    The contract for the design, construction and commissioning of the Replacement Research Reactor was signed in July 2000. This was followed by the completion of the detailed design and an application for a construction licence was made in May 2001. This paper will describe the main elements of the design and their relation to the proposed applications of the reactor. The future stages in the project leading to full operation are also described

  10. Advanced Burner Reactor Preliminary NEPA Data Study

    International Nuclear Information System (INIS)

    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly

  11. Advanced Burner Reactor Preliminary NEPA Data Study.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Cahalan, J. E.; Deitrich, L. W.; Fanning, T. H.; Grandy, C.; Kellogg, R.; Kim, T. K.; Yang, W. S.; Nuclear Engineering Division

    2007-10-15

    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly

  12. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

    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

  13. System-integrated modular advanced reactor (SMART)

    International Nuclear Information System (INIS)

    Since the Kori nuclear power plant unit 1, the first nuclear power plant unit ever dedicated in Korea, began commercial operations with a generating capacity of 587 MW in 1978, much research and development has been conducted in the nuclear industry. In the middle 1980s, the Korean standard nuclear power plant (KSNP) was first developed under the 'nuclear power promotion plan' promulgated by the government with reference to system 80 of ABBCE of the USA. Applying indigenously accumulated technologies and up-to-date design standards from both home and abroad, the initial KSNP project began with the construction of the Younggwang NPP units No. 3 and 4. In addition, the Korea Atomic Energy Research Institute (KAERI) designed and constructed a high performance multipurpose research reactor based on experience in the operation of previous reactors and accumulated nuclear technology. Timed with completion of construction in April 1995, the reactor was named HANARO (high-flux advanced neutron application reactor), which, in Korean means, 'uniqueness'. In the middle of the 1990s, research and development was launched related to small and medium sized reactors (SMRs) to promote the utilization of nuclear energy. SMRs are under development worldwide for various purposes such as district heating, seawater desalination, nuclear ship propulsion, as well as electricity production. Generally, modern SMRs for power generation are expected to have greater simplicity of design, economy of mass production, and reduced capital costs. Many SMRs also have advantages of reactor safety and economics by implementing advanced design concepts and technology. Since 1997, KAERI has been developing the system-integrated modular advanced reactor (SMART), an advanced integral pressurized water reactor (PWR). The SMART is a promising, advanced SMR and has an integral type reactor with a rated thermal power of 330 MW. All major primary components, such as reactor core, steam generator (SG), main

  14. Advanced robotic remote handling system for reactor dismantlement

    International Nuclear Information System (INIS)

    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)

  15. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    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)

  16. Advanced simulation for fast reactor design

    International Nuclear Information System (INIS)

    Full text: This talk broadly reviews recent research aimed at applying advanced simulation techniques specifically to fast neutron reactors. By advanced simulation we generally refer to attempts to do more science-based simulation - that is, to numerically solve the three-dimensional governing physical equations on fine scales and observe and study the holistic phenomena that emerge. In this way simulation is treated more akin to a traditional physical experiment, and can can be used both separately and in conjunction with physical experiments to develop more accurate predictive theories on reactor behavior. Many existing fast reactor modeling tools were developed for last generation's computational resources. They were built by engineers and physicists with deep physical insight - insight that both shaped and was informed by existing theory, and was underpinned by a vast repository of experimental data. Their general approach was to develop models that were tailored to varying degrees to the details of the reactor design, using free model parameters that were subsequently calibrated to match existing experimental data. The resulting codes were thus extremely useful for their specific purpose but highly limited in their predictive capability (neutronics to a lesser degree). They tended to represent more the state-of-the-art in our understanding rather than tools of exploration and innovation. Recently, a number of researchers have attempted to study the feasibility of solving more fundamental governing equations on realistic, three-dimensional geometries for different fast reactor sub-domains. This includes solving the Navier-Stokes equations for single-phase sodium flow (Direct Numerical Simulation, Large Eddie Simulation, and Reynolds Averaged Navier Stokes Equations) in the core, upper plenum, primary and intermediate loop, etc.; the non-homogenized transport equations at very fine group, angle, and energy discretization, and thermo-mechanical feedback based on

  17. 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

  18. Reactor Materials Research

    International Nuclear Information System (INIS)

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel

  19. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  20. 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.

  1. Instrumentation to Enhance Advanced Test Reactor Irradiations

    International Nuclear Information System (INIS)

    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.

  2. Advanced fuels for fast reactors

    International Nuclear Information System (INIS)

    Full text: In addition to traditional fast reactor fuels that contain Uranium and Plutonium, the advanced fast reactor fuels are likely to include the minor actinides [Neptunium (Np), Americium (Am) and Curium (Cm)]. Such fuels are also referred to as transmutation fuels. The goal of transmutation fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a traditional fast spectrum nuclear fuel while destroying recycled actinides. Oxide, metal, nitride, and carbide fuels are candidates under consideration for this application, based on historical knowledge of fast reactor fuel development and specific fuel tests currently being conducted in international transmutation fuel development programs. Early fast reactor developers originally favored metal alloy fuel due to its high density and potential for breeder operation. The focus of pressurized water reactor development on oxide fuel and the subsequent adoption by the commercial nuclear power industry, however, along with early issues with low burnup potential of metal fuel (now resolved), led later fast reactor development programs to favor oxide fuels. Carbide and nitride fuels have also been investigated but are at a much lower state of development than metal and oxide fuels, with limited large scale reactor irradiation experience. Experience with both metal and oxide fuels has established that either fuel type will meet performance and reliability goals for a plutonium fueled fast spectrum test reactor, both demonstrating burnup capability of up to 20 at.% under normal operating conditions, when clad with modified austenitic or ferritic martensitic stainless steel alloys. Both metal and oxide fuels have been shown to exhibit sufficient margin to failure under transient conditions for successful reactor operation. Summary of selected fuel material properties taken are provided in the paper. The main challenge for the development of transmutation fast reactor

  3. Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor, United States of America

    International Nuclear Information System (INIS)

    The aging of research reactors worldwide has resulted in a heightened awareness in the international decommissioning community of the timeliness to review and address the needs of research reactor operators in planning for and eventually performing the decommissioning of these types of facilities. Many reactors already undergoing decommissioning can be used as test beds for evaluating enhanced or new/innovative technologies for decommissioning; it is possible that new techniques could be made available for future research reactor-decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the decommissioners in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to their use in actual research reactor decommissioning. The decommissioning of the CP-5 Research Reactor located at the ANL-East Site has been completed. In this paper we present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors. In addition, details are provided on other related U.S. D and D activities, which may be useful to the international research reactor D and D community. (author)

  4. Research for enhancing reactor safety

    International Nuclear Information System (INIS)

    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)

  5. Industrial structure at research reactor suppliers

    International Nuclear Information System (INIS)

    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

  6. Gaseous fuel reactor research

    Science.gov (United States)

    Thom, K.; Schneider, R. T.

    1977-01-01

    The paper reviews studies dealing with the concept of a gaseous fuel reactor and describes the structure and plans of the current NASA research program of experiments on uranium hexafluoride systems and uranium plasma systems. Results of research into the basic properties of uranium plasmas and fissioning gases are reported. The nuclear pumped laser is described, and the main results of experiments with these devices are summarized.

  7. The current status of utilization of research reactors in China

    International Nuclear Information System (INIS)

    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)

  8. Role of advanced reactors for sustainable development

    International Nuclear Information System (INIS)

    Availability of energy is an important prerequisite for socio-economic development in all parts of the world. Nuclear energy is an essentially unlimited energy resource with the potential to provide 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 national 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 plants. This experience forms a sound basis for further improvements. Nuclear programmes in 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 nuclear power programmes and offers assistance to countries with an interest in exploratory or research programmes. The paper gives an overview about trends in advanced reactor development programmes in Member States and the possible application of advanced reactors for electricity generation and heat production. (author)

  9. Advanced methods in teaching reactor physics

    International Nuclear Information System (INIS)

    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.

  10. Utilization of nuclear research reactors

    International Nuclear Information System (INIS)

    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)

  11. PROTEUS research reactor

    International Nuclear Information System (INIS)

    The PROTEUS zero power reactor at the Paul Scherrer Institute (PSI) in Switzerland achieved first criticality in 1968 and since then has been operated as an experimental tool for reactor physics research on test lattices representative of a wide range of reactor concepts. Reactor design codes and their associated data libraries are validated on the basis of the experimental results obtained. PROTEUS is normally configured as a driven system, in which a subcritical test zone is made critical by the surrounding driver zones. The advantages of driven systems can be summarized as follows: - Smaller amount of test fuel is required; - Large range of test zone conditions (including k∞ < 1 states) can be investigated by changes in the driver loading alone, thus avoiding undesirable perturbations to the test zone which would influence the measurement conditions and thus affect the interpretability of the results; - Necessary reactor control and instrumentation equipment (usually perturbing from the experimental viewpoint) can be located in the outer driver regions, thereby avoiding disturbance of the test lattice

  12. 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.

  13. Accident analysis in research reactors

    International Nuclear Information System (INIS)

    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)

  14. 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.

  15. Technical Meeting (Research Coordination Meeting) 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)

    The Technical Meeting held at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) in Hefei was the second Research Coordination Meeting (RCM) of the CRP on 'Studies of Innovative Reactor Technology Options for Effective Incineration of Radioactive Waste'. All but one Member States participating in the CRP were attending: in all, 26 participants from 13 Member States and three international organizations. The overall objective of the CRP is to increase the capability of Member States in developing and applying advanced nuclear technologies in the area of long-lived radioactive waste utilization and transmutation. The final goal of the CRP is to deepen the understanding of the dynamics of transmutation systems, e.g., the accelerator driven system, especially of systems with deteriorated safety parameters, to qualify the available methods, specify the range of validity of these methods, and formulate requirements for future theoretical developments. Should transient experiments be available, the CRP will pursue experimental benchmarking work. Based on the results, the CRP will conclude on the potential need of transient experiments and make appropriate proposals for experimental programs. The scope of the second RCM was to review the progress achieved with regard on the technical work of the CRP. In particular, the main objectives of the RCM were to (a) analyse and inter-compare the individual results; (b) identify of eventual changes/improvements to the tasks and/or work plans; (c) plan the next stage(s); and (d) start preparation of the final CRP report. The participants were given a brief overview of the Institute's mission and accomplishments. ASIPP (Academia Sinica, Institute of Plasma Physics) was founded in 1978 as the leading centre for high temperature plasma physics, magnetically confined fusion technology, as well as R and D in related technological areas. ASIPP employs nearly 500 staff, of which more than 70% are scientists or engineers

  16. Report on the workshop "Decay spectroscopy at CARIBU: advanced fuel cycle applications, nuclear structure and astrophysics". 14-16 April 2011, Argonne National Laboratory, USA.

    Energy Technology Data Exchange (ETDEWEB)

    Kondev, F.; Carpenter, M.P.; Chowdhury, P.; Clark, J.A.; Lister, C.J.; Nichols, A.L.; Swewryniak, D. (Nuclear Engineering Division); (Univ. of Massachusetts); (Univ. of Surrey)

    2011-10-06

    A workshop on 'Decay Spectroscopy at CARIBU: Advanced Fuel Cycle Applications, Nuclear Structure and Astrophysics' will be held at Argonne National Laboratory on April 14-16, 2011. The aim of the workshop is to discuss opportunities for decay studies at the Californium Rare Isotope Breeder Upgrade (CARIBU) of the ATLAS facility with emphasis on advanced fuel cycle (AFC) applications, nuclear structure and astrophysics research. The workshop will consist of review and contributed talks. Presentations by members of the local groups, outlining the status of relevant in-house projects and availabile equipment, will also be organized. time will also be set aside to discuss and develop working collaborations for future decay studies at CARIBU. Topics of interest include: (1) Decay data of relevance to AFC applications with emphasis on reactor decay heat; (2) Discrete high-resolution gamma-ray spectroscopy following radioactive decya and related topics; (3) Calorimetric studies of neutron-rich fission framgents using Total ABsorption Gamma-Ray Spectrometry (TAGS) technique; (4) Beta-delayed neutron emissions and related topics; and (5) Decay data needs for nuclear astrophysics.

  17. Fast breeder reactor research

    International Nuclear Information System (INIS)

    reactors of the future, the body of research aimed at developing liquid metal cooled fast reactors, national plans for work in 1976 on developing fast reactors - these were some of the topics discussed in connection with the national programmes. The development of power reactors involves a wide range of problems in the fields of nuclear and reactor physics, the thermophysics, chemistry, physics and technology of the cooling system, structural materials and nuclear fuel, the fabrication of reliable fuel elements and operating equipment, reactor monitoring and control, spent fuel reprocessing, the economics of constructing fast power reactors, nuclear safety, etc. The IWGFR, as at previous meetings, therefore paid great attention to the matter of holding international specialists' meetings. The working group recommended that the IAEA should organize the following IWGFR meetings in 1976: (1) In-Service Inspection and Monitoring (Bensberg, FRG, March 1976). (2) Cavitation in Sodium and Studies of Analogy with Water as Compared to Sodium (Cadarache, France, April 1976). (3) High Temperature Structural Design Technology (United States, May 1976) (4) Aerosol Formation, Vapour Deposits and Sodium Vapour Trapping (France, September-December 1976). The Group welcomed the IAEA's proposal to hold specialists' meetings on 'Fast Reactor Instrumentation' and 'Fuel Reprocessing and Recycling Techniques' within the framework of the Agency's programme of working groups in 1976. After discussing questions of co-ordinating and organizing international conferences on fast reactors, the IWGFR agreed to send representatives to the joint meeting of the American Nuclear Society and the American Institute of Metallurgical Engineers on 'Liquid Metal Technology', to be held at Champion, Pennsylvania, U.S.A. from 3-6 May 1976, and recommended that the IAEA should organize an international symposium on the 'Design, Construction and Operating Experience of Demonstration Fast Power Reactors' at Bologna

  18. The concept of a research fusion reactor

    International Nuclear Information System (INIS)

    Thus,for advancement towards a commercial fusion reactor,we have proposed here as a next step a steady state operated research fusion reactor with an increased plasma-wall detachment so as to further guarantee not only the production but also a long-term (for many years) confinement of a self-sustained plasma at the existing technology level. We consider the primary goal of the research fusion reactor is the provision of full-scale conditions for carrying out materials science experiments to create and test 1 st wall materials for the commercial fusion reactor

  19. MIT research reactor. Power uprate and utilization

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

  1. 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.

  2. The Advanced Light Water Reactor

    International Nuclear Information System (INIS)

    The U. S. Advanced Light Water Reactor Program is a forward-looking program designed to produce viable nuclear generating system candidates to meet the very real, and perhaps imminent, need for new power generation capacity in the U. S. and around the world. The ALRR Program is an opportunity to move ahead with confidence, to confront problems today which must be confronted if the U. S. electrical utilities are to continue to meet their commitment to provide safe, reliable, economical electrical power to the nation in the years ahead. Light water reactor technology is today playing a vital role in the production of electricity to meet the world's needs. At present about 13% of the world's electricity is supplied by nuclear power plants, most of those light water reactors. Nevertheless, there is a clear need for expanded use of nuclear generation. Here in Korea and elsewhere in Asia, demand for electricity has continued to increase at a very high rate. In the United States demand growth has been more moderate, but a large number of existing stations will be ready for replacement in the next two decades, and all countries face the problem of dwindling fuel supplies and growing environmental impact of fossil-fired power plants. Despite the evident need for expanded nuclear generation capacity in the United States, there have been no new plants ordered in the past ten years and at present there are no immediate prospects for new plant orders. Concerns about safety, the high cost of recent nuclear stations, and the current excess of electrical generation capacity in the United States, have combined to interrupt completely the growth of this vital power supply system

  3. Research Reactors of Ukraine

    International Nuclear Information System (INIS)

    Ukraine today operates two nuclear research reactors: WWR-M (total capacity of 10 MW), which is located on the site of the Kyiv Nuclear Research Institute of the National Academy of Sciences of Ukraine, and IR-100 (total capacity of 200 kW), which is located on the site of Sevastopol National University of Nuclear Energy and Industry. Both of them have been in operation since the 1960s. The operation project period of WWR-M for which it is licensed is limited to 31 December 2013. In order to improve safety at WWR-M several modernization projects, development of the reactor vessel and the first loop equipment ageing management programme were conducted. According to the license for operation of IR-100 the operation period of the reactor depends upon results from assessments of critical safety elements such as the tank, control and protection system, cable lines and electrical switchgear. Currently the operation period of this equipment has been justified until 2013. (author)

  4. Advanced gas-cooled reactors (AGR)

    International Nuclear Information System (INIS)

    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

  5. Decommissioning of research reactors

    International Nuclear Information System (INIS)

    Research reactors of WWR-S type were built in countries under Soviet influence in '60, last century and consequently reached their service life. Decommissioning implies removal of all radioactive components, processing, conditioning and final disposal in full safety of all sources on site of radiological pollution. The WWR-S reactor at Bucuresti-Magurele was put into function in 1957 and operated until 1997 when it was stopped and put into conservation in view of decommissioning. Presented are three decommissioning variants: 1. Reactor shut-down for a long period (30-50 years) what would entail a substantial decrease of contamination with lower costs in dismantling, mechanical, chemical and physical processing followed by final disposal of the radioactive wastes. The drawback of this solution is the life prolongation of a non-productive nuclear unit requiring funds for personnel, control, maintenance, etc; 2. Decommissioning in a single stage what implies large funds for a immediate investment; 3. Extending the operation on a series of stages rather phased in time to allow a more convenient flow of funds and also to gather technical solutions, better than the present ones. This latter option seems to be optimal for the case of the WWR-S Research at Bucharest-Magurele Reactor. Equipment and technologies should be developed in order to ensure the technical background of the first operations of decommissioning: equipment for scarification, dismantling, dismemberment in a highly radioactive environment; cutting-to-pieces and disassembling technologies; decontamination modern technologies. Concomitantly, nuclear safety and quality assurance regulations and programmes, specific to decommissioning projects should be implemented, as well as a modern, coherent and reliable system of data acquisition, recording and storing. Also the impact of decommissioning must be thoroughly evaluated. The national team of specialists will be assisted by IAEA experts to ensure the

  6. Advanced test reactor. Testing capabilities and plans

    International Nuclear Information System (INIS)

    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. For future research, some ATR modifications and enhancements are currently planned. In 2007 the US Department of Energy designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR for material testing research by a broader user community. This paper provides more details on some of the ATR capabilities, key design features, experiments, and plants for the NSUF. (author)

  7. A premature demise for RERTR [Reduced Enrichment for Research and Test Reactors programme]?

    International Nuclear Information System (INIS)

    A common commitment from France, Belgium, Germany and the US to eliminate highly enriched uranium from their research reactors is needed to help guard against this material falling into the wrong hands. In the US, an essential part of this commitment would be rekindling the weakened Reduced Enrichment for Research and Test Reactors programme (RERTR). This is an American initiative to develop low-enrichment uranium fuel for research reactors that have previously required weapons-usable material. Underway since 1978 at Argonne National Laboratory, RERTR has achieved some impressive results: the development of higher density, low enriched fuels that are suitable for use at over 90% of the world's research reactors; a net reduction of US exports of highly enriched uranium (HEU) from the annual 700kg levels in the late 1970s to a 1990 level of just over 100kg; the encouragement of international scientific co-operation aimed at developing new fuels and facilitating the conversion of existing reactors to these fuels. However, in recent years, the US commitment to RERTR has been declining -budgets have fallen and advanced fuel development work has terminated. (author)

  8. Advances in laser solenoid fusion reactor design

    International Nuclear Information System (INIS)

    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

  9. Modern research reactors in the world and RA research reactor

    International Nuclear Information System (INIS)

    This paper covers the following topics: fundamentals of research reactors, thermal neutron flux density, classification of research reactors in the world, properties of research reactors of higher power in the world according to IAEA data for 1995, their application, and trend of development, experimental feasibility and status of RA reactor. Trend of research reactors development in the world (after 1980) is directed towards increasing the neutron production quality factor, i.e. ratio between thermal neutron flux density and reactor power, which is achieved by designing compact reactor cores. With the aim of renewal of RA reactor (without analysis of reactor components and staff aging, possibility of restart and commercialization), according to the analysis in this paper, it can be concluded: there is very few reactors under construction in the world, all the important countries in Europe have research reactors; RA reactor is not very interesting for development of reactor physics; nowadays RA reactor is in the group of reactors which are 30-40 years old; its inventories of fuel and heavy water are enough for about 20 years of operation; it has achieved high quality factor of neutron production with low and highly enriched fuel; core transfer from low highly enriched to low enriched fuel should be carefully studies from operation, experimental and economical point of view; it is necessary to use the advantages of RA reactor (minimum investment): volume of the core and reflector which enables availability of neutron flux for the users (numerous experimental loops), fuel in shape of slugs enabling efficient fuel management and flexible neutron flux distribution in the core in the reflector, reactor operation should be directed towards commercial applications. Bibliography of more than 140 relevant papers used is included in this paper

  10. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    In 1998, the NNSA organized to complete the nuclear safety review on the test loop in-reactor operation of the High-flux Engineering Experimental Reactor (HFEER) and the re-operation of the China Pulsed Reactor and the Uranium-water Criticality Facility. The NNSA conducted the nuclear safety review on the CP application of the China Experimental Fast Reactor (CEFR) and the siting of China Advanced Research Reactor (CARR), and carried out the construction supervision on HTR-10, and dealt with the event about the technological tube breakage of HWRR and other events

  11. Coordinated research project (CRP) on studies of advanced reactor technology options for effective incineration of radioactive waste - Scope and objectives

    International Nuclear Information System (INIS)

    The overall objective of the CRP is to perform R and D tasks contributing towards the proof of practicality for long-lived waste transmutation. For a sound assessment of the transient and accident behaviour, the neutron kinetics and dynamics have to be qualified, especially as the margins for the safety relevant neutronics parameters are becoming small in a 'dedicated' transmuter. The CRP will integrate benchmarking of transient/accident simulation codes focussing on the phenomena and effects relevant to various critical and sub-critical systems under severe neutron flux changes and rearrangements. The CRP will investigate future needs both for theoretical means (data, codes) and experimental information related to the various transmutation systems. The final goals of the CRP are to (a) deepen the understanding of the dynamics of transmutation systems, e.g., the accelerator driven system, especially systems with deteriorated safety parameters, (b) qualify the available methods and specify their range of validity, and (c) formulate requirements for future theoretical developments. Should transient experiments be available, the CRP might also 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 CRP will consider various transmuter and actinides incinerator concepts, from traditional to very advanced, both critical and sub-critical. No optimization will be performed, experience and good engineering judgment is used to define the design parameters of the various concepts. The generic dynamic behavior of the different systems will be assessed and inter-comparisons will be performed. The concepts are derived from those proposed by the participants in this RCM. Later additions might be considered. The studies will include static, kinetics and dynamics calculations. Influence of burnup on transient behavior will be

  12. Design of a multipurpose research reactor

    International Nuclear Information System (INIS)

    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

  13. Mimic of OSU research reactor

    International Nuclear Information System (INIS)

    The Ohio State University research reactor (OSURR) is undergoing improvements in its research and educational capabilities. A computer-based digital data acquisition system, including a reactor system mimic, will be installed as part of these improvements. The system will monitor the reactor system parameters available to the reactor operator either in digital parameters available to the reactor operator either in digital or analog form. The system includes two computers. All the signals are sent to computer 1, which processes the data and sends the data through a serial port to computer 2 with a video graphics array VGA monitor, which is utilized to display the mimic system of the reactor

  14. MINT research reactor safety program

    Energy Technology Data Exchange (ETDEWEB)

    Mohamad Idris bin Taib [Division of Special Project, Malaysian Institute for Nuclear Technology Research (MINT), Bangi (Malaysia)

    2000-11-01

    Malaysian Institute for Nuclear Technology Research (MINT) Research Reactor Safety Program has been done along with Reactor Power Upgrading Project, Reactor Safety Upgrading Project and Development of Expert System for On-Line Nuclear Process Control Project. From 1993 up to date, Neutronic and Thermal-hydraulics analysis, Probabilistic Safety Assessment as well as installation of New 2 MW Secondary Cooling System were done. Installations of New Reactor Building Ventilation System, Reactor Monitoring System, Updating of Safety Analysis Report and Upgrading Primary Cooling System are in progress. For future activities, Reactor Modeling will be included to add present activities. (author)

  15. The advanced liquid metal reactor actinide recycle system

    International Nuclear Information System (INIS)

    The current U.S. National Energy Strategy includes four key goals for nuclear policy: enhance safety and design standards, reduce economic risk, reduce regulatory risk, and establish an effective high-level nuclear waste program. The U.S. Department of Energy's Advanced Liquid Metal Reactor Actinide Recycle System is consistent with these objectives. The system has the ability to fulfill multiple missions with the same basic design concept. In addition to providing an option for long-term energy security, the system can be effectively utilized for recycling of actinides in light water reactor (LWR) spent fuel, provide waste management flexibility, including the reduction in the waste quantity and storage time and utilization of the available energy potential of LWR spent fuel. The actinide recycle system is comprised of (1) a compact liquid metal (sodium) cooled reactor system with optimized passive safety characteristics, and (2) pyrometallurgical metal fuel cycle presently under development of Argonne National Laboratory. The waste reduction of LWR spent fuel is accomplished by transmutation or fissioning of the longer-lived transuranic isotopes to shorter-lived fission products in the reactor. In this presentation the economical and environmental incentive of the actinide recycle system is addressed and the status of development including licensing aspects is described. 3 refs., 1 tab., 6 figs

  16. Preliminary design concept of an advanced integral reactor

    International Nuclear Information System (INIS)

    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 reactor 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)

  17. Penn State advanced light water reactor concept

    International Nuclear Information System (INIS)

    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

  18. Advanced light and heavy water reactors for improved fuel utilization

    International Nuclear Information System (INIS)

    On 26-29 November 1984 the Agency convened at its Headquarters in Vienna the Technical Committee and Workshop on Advanced Light and Heavy Water Reactor Technology in order to provide an opportunity to review and discuss the current status and recent development in the lay-out and design of advanced water reactor and to identify areas in which additional research and development are needed. The meeting was attended by 45 participants from 16 nations and 2 international organizations presenting 25 papers. The Conference presentations were divided into sessions devoted to the following topics: Advanced light water reactor programmes (6 papers); Advanced light water design, technology and physics (12 papers); Advanced heavy water reactors (7 papers). A separate abstract was prepared for each of these papers

  19. Advanced CANDU reactor, evolution and innovation

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has developed the ACRTM (Advanced CANDU(1) ReactorTM) to meet today's market challenges. It is a light water tube type pressurized water reactor and is the latest evolution of CANDU technology. The design was launched to be cost-competitive with other generating sources, while building on the unique safety and operational advantages of the CANDU design. The ACR is an evolutionary design that retains the proven CANDU features delivered at Qinshan Phase III, while incorporating a set of innovative features and proven state-of-the-art technologies that have emerged from AECL's ongoing Research, Development and Demonstration programs. This approach ensures that key design parameters are well supported by existing reactor experience and R and D. The result is a design that delivers a new threshold in safety, performance and economics while retaining ample design margin. AECL has developed the enabling technologies and components for the ACR design, and has applied them to two plant sizes, ACR-700 and ACR-1000. The ACR integrates hallmark characteristics of traditional CANDU plants (e.g. horizontal pressure tubes, on power fuelling, automated reactor control systems, and dual independent shutdown systems), new innovations (e.g. state-of-the- art control room, extensive use of modular construction techniques, smaller reactor core, enriched uranium fuel), and certain PWR features (e.g. light water coolant, negative void reactivity). The ACR is designed for a high capacity factor and low operation and maintenance costs. It fully exploits the construction techniques that contributed to the impressive schedule accomplishments at Qinshan Phase III and therefore features a very short construction schedule, 40 months construction schedule (First Concrete to Fuel Loading ) for the first unit with improvements to 36 months for later units. The ACR is a true Gen-III plus product with a broad application. It has been proven to be an ideal

  20. 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)

    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

  1. Argonne National Laboratory research to help U.S. steel industry

    CERN Multimedia

    2003-01-01

    Argonne National Laboratory has joined a $1.29 million project to develop technology software that will use advanced computational fluid dynamics (CFD), a method of solving fluid flow and heat transfer problems. This technology allows engineers to evaluate and predict erosion patterns within blast furnaces (1 page).

  2. 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 Sandia's 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.

  3. Conceptual design of the advanced marine reactor MRX

    Science.gov (United States)

    1991-02-01

    Design studies on the advanced marine reactors have been done continuously since 1983 at the Japan Atomic Energy Research Institute (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 pressurized water reactor (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.

  4. Decontamination and decommissioning of the Experimental Boiling Water Reactor at Argonne National Laboratory

    International Nuclear Information System (INIS)

    The Experimental Boiling Water Reactor (EBWR), located on the Argonne National Laboratory-East (ANL-E) site, started operations in 1957. The initial rating was 20 MW(t). The rating was eventually increased to 70 MW(t) in 1959 and 100 MW(t) in 1962. The reactor was shut down in 1967 and all of the fuel was removed from the facility. The facility was placed in dry lay-up until 1986. ANL-E personnel started the decontamination and decommissioning (D ampersand D) effort in 1986. Supporting equipment such as the external steam system and some of the upper reactor components, the core riser and the top fuel shroud, were removed at that time. Characterization of the facility was also undertaken. The contract to complete the EBWR D ampersand D Project was issued in December 1993. The initial schedule called for the final effort to be divided into five phases that were to be completed over a four year period. However, this schedule was subsequently consolidated, at the request of ANL-E, to a thirteen month period, with the on-site work to be completed by the end of 1994. The EBWR D ampersand D Project is approximately 88% complete. A small quantity of reactor internals remains to be volume reduced along with the removal of the SFSP water treatment system. Upon completion of this work the facility will be decontaminated and a final survey completed. The planned completion of on-site work is scheduled for July 1995

  5. 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)

  6. Physical security at research reactors

    International Nuclear Information System (INIS)

    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

  7. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided 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.

  8. NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

    International Nuclear Information System (INIS)

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided 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

  9. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided 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.

  10. NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

    International Nuclear Information System (INIS)

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided 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

  11. Light-Water-Reactor safety research program. Quarterly progress report, January--March 1977

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-06-01

    The report summarizes the Argonne National Laboratory work performed during January, February, and March 1977 on water-reactor-safety problems. The following research and development areas are covered: (1) loss-of-coolant accident research: heat transfer and fluid dynamics; (2) transient fuel response and fission-product release program; (3) mechanical properties of zircaloy containing oxygen; and (4) steam-explosion studies.

  12. Development of essential system technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    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)

  13. Development of essential system technologies for advanced reactor

    International Nuclear Information System (INIS)

    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)

  14. Status of advanced technologies for CANDU reactors

    International Nuclear Information System (INIS)

    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

  15. Relevant thermal hydraulic aspects of advanced reactors design: status report

    International Nuclear Information System (INIS)

    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

  16. Conceptual design of multipurpose compact research reactor

    International Nuclear Information System (INIS)

    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)

  17. Management of research reactor ageing

    International Nuclear Information System (INIS)

    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

  18. Research reactors and alternative devices for research

    International Nuclear Information System (INIS)

    This report includes papers on research reactors and alternatives to the research reactors - radioisotopic neutron sources, cyclotrons, D-T neutron generators and small accelerators, used for radioisotope production, neutron activation analysis, material science, applied and basic research using neutron beams. A separate abstract was prepared for each of the 7 papers

  19. Argonne National Laboratory, High Energy Physics Division, semiannual report of research activities, July 1, 1989--December 31, 1989

    Energy Technology Data Exchange (ETDEWEB)

    1989-01-01

    This report discusses research being conducted at the Argonne National Laboratory in the following areas: Experimental High Energy Physics; Theoretical High Energy Physics; Experimental Facilities Research; Accelerator Research and Development; and SSC Detector Research and Development.

  20. Argonne National Laboratory, High Energy Physics Division, semiannual report of research activities, July 1, 1989--December 31, 1989

    International Nuclear Information System (INIS)

    This report discusses research being conducted at the Argonne National Laboratory in the following areas: Experimental High Energy Physics; Theoretical High Energy Physics; Experimental Facilities Research; Accelerator Research and Development; and SSC Detector Research and Development

  1. Structural materials challenges for advanced reactor systems

    Science.gov (United States)

    Yvon, P.; Carré, F.

    2009-03-01

    Key technologies for advanced nuclear systems encompass high temperature structural materials, fast neutron resistant core materials, and specific reactor and power conversion technologies (intermediate heat exchanger, turbo-machinery, high temperature electrolytic or thermo-chemical water splitting processes, etc.). The main requirements for the materials to be used in these reactor systems are dimensional stability under irradiation, whether under stress (irradiation creep or relaxation) or without stress (swelling, growth), an acceptable evolution under ageing of the mechanical properties (tensile strength, ductility, creep resistance, fracture toughness, resilience) and a good behavior in corrosive environments (reactor coolant or process fluid). Other criteria for the materials are their cost to fabricate and to assemble, and their composition could be optimized in order for instance to present low-activation (or rapid desactivation) features which facilitate maintenance and disposal. These requirements have to be met under normal operating conditions, as well as in incidental and accidental conditions. These challenging requirements imply that in most cases, the use of conventional nuclear materials is excluded, even after optimization and a new range of materials has to be developed and qualified for nuclear use. This paper gives a brief overview of various materials that are essential to establish advanced systems feasibility and performance for in pile and out of pile applications, such as ferritic/martensitic steels (9-12% Cr), nickel based alloys (Haynes 230, Inconel 617, etc.), oxide dispersion strengthened ferritic/martensitic steels, and ceramics (SiC, TiC, etc.). This article gives also an insight into the various natures of R&D needed on advanced materials, including fundamental research to investigate basic physical and chemical phenomena occurring in normal and accidental operating conditions, lab-scale tests to characterize candidate materials

  2. Applied mathematical sciences research at Argonne, April 1, 1981-March 31, 1982

    International Nuclear Information System (INIS)

    This report reviews the research activities in Applied Mathematical Sciences at Argonne National Laboratory for the period April 1, 1981, through March 31, 1982. The body of the report discusses various projects carried out in three major areas of research: applied analysis, computational mathematics, and software engineering. Information on section staff, visitors, workshops, and seminars is found in the appendices

  3. Applications of the advanced neutron source reactor

    International Nuclear Information System (INIS)

    When the technique of neutron scattering was pioneered at the X-10 graphite reactor at Oak Ridge National Laboratory about 50 years ago, it was used to study certain important, but fairly esoteric, properties of crystals. From this modest beginning, neutron scattering has become a major tool in every branch of science, from the astrophysics of the early universe to human biology, and in many important industrial and engineering applications. In a typical modern research reactor it is not unusual to find one instrument studying new polymeric materials, while its neighbor is measuring residual stress in a jet turbine, sometimes with the jet operating. Most of this development has taken place outside of the United States, primarily in Western Europe, Japan and Russia, and it is generally recognized that we are a decade behind our competitors in this important field. The Advanced Neutron Source (ANS), planned to become operational as a user-facility at Oak Ridge at the end of this decade, will regain our leadership in neutron-based research and will be a major center for attracting new students into science. This paper discusses some of the research and development applications of the ANS, with an emphasis on applied materials science and engineering

  4. Fusion reactor research

    International Nuclear Information System (INIS)

    This work covers four separate areas: (1) development of technology for processing liquid lithium from blankets, (2) investigation of hydrogen isotope permeation in candidate structural metals and alloys for near-term fusion reactors, (3) analytical studies encompassing fusion reactor thermal hydraulics, tritium facility design, and fusion reactor safety, and (4) studies involving dosimetry and damage analysis. Recent accomplishments in each of these areas are summarized

  5. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    International Nuclear Information System (INIS)

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting

  6. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  7. Improvement of research reactor sustainability

    International Nuclear Information System (INIS)

    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)

  8. Preliminary design concepts of an advanced integral reactor

    International Nuclear Information System (INIS)

    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

  9. Introduction of advanced pressurized water reactors in France

    International Nuclear Information System (INIS)

    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

  10. Analytical chemistry requirements for advanced reactors

    International Nuclear Information System (INIS)

    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

  11. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm3 was by then in routine use, illustrated how far work has progressed

  12. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

  13. Advanced power reactors with improved safety characteristics

    International Nuclear Information System (INIS)

    The primary objective of nuclear safety is the protection of individuals, society and environment against radiological hazards from accidental releases of radioactive materials contained in nuclear reactors. Hereto, these materials are enclosed by several successive barriers and the barriers protected against mishaps and accidents by a multi-level system of safety precautions. The evolution of reactor technology continuously improves this concept and its implementation. At a world-wide scale, several advanced reactor concepts are currently being considered, some of them already at a design stage. Essential safety objectives include both further strengthening the prevention of accidents and improving the containment of fission products should an accident occur. The proposed solutions differ considerably with regard to technical principles, plant size and time scales considered for industrial application. Two typical approaches can be distinguished: The first approach basically aims at an evolution of power reactors currently in use, taking into account the findings from safety research and from operation of current plants. This approach makes maximum use of proven technology and operating experience but may nevertheless include new safety features. The corresponding designs are often termed 'large evolutionary'. The second approach consists in more fundamental changes compared to present designs, often with strong emphasis on specific passive features protecting the fuel and fuel cladding barriers. Owing to the nature and capability of those passive features such 'innovative designs' are mostly smaller in power output. The paper describes the basic objectives of such developments and illustrates important technical concepts focusing on next generation plants, i.e. designs to be available for industrial application until the end of this decade. 1 tab. (author)

  14. Argonne National Laboratory: Laboratory Directed Research and Development FY 1993 program activities. Annual report

    Energy Technology Data Exchange (ETDEWEB)

    None

    1993-12-23

    The purposes of Argonne`s Laboratory Directed Research and Development (LDRD) Program are to encourage the development of novel concepts, enhance the Laboratory`s R&D capabilities, and further the development of its strategic initiatives. Projects are selected from proposals for creative and innovative R&D studies which are not yet eligible for timely support through normal programmatic channels. Among the aims of the projects supported by the Program are establishment of engineering ``proof-of-principle`` assessment of design feasibility for prospective facilities; development of an instrumental prototype, method, or system; or discovery in fundamental science. Several of these projects are closely associated with major strategic thrusts of the Laboratory as described in Argonne`s Five Year Institutional Plan, although the scientific implications of the achieved results extend well beyond Laboratory plans and objectives. The projects supported by the Program are distributed across the major programmatic areas at Argonne as indicated in the Laboratory LDRD Plan for FY 1993.

  15. Decontamination and decommissioning of the Argonne Thermal Source Reactor at Argonne National Laboratory - East project final report.

    Energy Technology Data Exchange (ETDEWEB)

    Fellhauer, C.; Garlock, G.; Mathiesen, J.

    1998-12-02

    The ATSR D&D Project was directed toward the following goals: (1) Removal of radioactive and hazardous materials associated with the ATSR Reactor facility; (2) Decontamination of the ATSR Reactor facility to unrestricted use levels; and (3)Documentation of all project activities affecting quality (i.e., waste packaging, instrument calibration, audit results, and personnel exposure). These goals had been set in order to eliminate the radiological and hazardous safety concerns inherent in the ATSR Reactor facility and to allow, upon completion of the project, unescorted and unmonitored access to the area. The reactor aluminum, reactor lead, graphite piles in room E-111, and the contaminated concrete in room E-102 were the primary areas of concern. NES, Incorporated (Danbury, CT) characterized the ATSR Reactor facility from January to March 1998. The characterization identified a total of thirteen radionuclides, with a total activity of 64.84 mCi (2.4 GBq). The primary radionuclides of concern were Co{sup 60}, Eu{sup 152}, Cs{sup 137}, and U{sup 238}. No additional radionuclides were identified during the D&D of the facility. The highest dose rates observed during the project were associated with the reactor tank and shield tank. Contact radiation levels of 30 mrem/hr (0.3 mSv/hr) were measured on reactor internals during dismantlement of the reactor. A level of 3 mrem/hr (0.03 mSv/hr) was observed in a small area (hot spot) in room E-102. DOE Order 5480.2A establishes the maximum whole body exposure for occupational workers at 5 rem/yr (50 mSv/yr); the administrative limit at ANL-E is 1 rem/yr (10 mSv/yr).

  16. 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.

  17. Advanced nuclear reactor types and technologies

    International Nuclear Information System (INIS)

    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

  18. Argonne National Laboratory High Energy Physics Division semiannual report of research activities, January 1, 1989--June 30, 1989

    Energy Technology Data Exchange (ETDEWEB)

    1989-01-01

    This paper discuss the following areas on High Energy Physics at Argonne National Laboratory: experimental program; theory program; experimental facilities research; accelerator research and development; and SSC detector research and development.

  19. Argonne National Laboratory High Energy Physics Division semiannual report of research activities, January 1, 1989--June 30, 1989

    International Nuclear Information System (INIS)

    This paper discuss the following areas on High Energy Physics at Argonne National Laboratory: experimental program; theory program; experimental facilities research; accelerator research and development; and SSC detector research and development

  20. Replacement research reactor for Australia

    International Nuclear Information System (INIS)

    In 1992, the Australian Government commissioned a review into the need for a replacement research reactor. That review concluded that in about years, if certain conditions were met, the Government could make a decision in favour of a replacement reactor. A major milestone was achieved when, on 3 September 1997, the Australian Government announced the construction of a replacement research reactor at the site of Australia's existing research reactor HIFAR, subject to the satisfactory outcome of an environmental assessment process. The reactor will be have the dual purpose of providing a first class facility for neutron beam research as well as providing irradiation facilities for both medical isotope production and commercial irradiations. The project is scheduled for completion before the end of 2005. (author)

  1. The research reactor TRIGA Mainz

    International Nuclear Information System (INIS)

    Paper dwells upon the design and the operation of one of the German test reactors, namely, the TRIGA Mainz one (TRIGA: Training Research Isotope Production General Atomic). The TRIGA reactor is a pool test reactor the core of which contains a graphite reflector and is placed into 2 m diameter and 6.25 m height aluminum vessel. There are 75 fuel elements in the reactor core, and any of them contains about 36 g of 235U. The TRIGA reactors under the stable operation enjoy wide application to ensure tests and irradiation, namely: neutron activation analysis, radioisotope production, application of a neutron beam to ensure the physical, the chemical and the medical research efforts. Paper presents the reactor basic experimental program lines

  2. Code qualification of structural materials for AFCI advanced recycling reactors

    International Nuclear Information System (INIS)

    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) and the

  3. 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

  4. The Ongkharak Nuclear Research Center (ONRC) research reactor project: a status review

    International Nuclear Information System (INIS)

    The new Ongkharak Nuclear Research Center in the vicinity of Bangkok, Thailand is planned to replace the more than 30 years old facilities located in the Chatuchak district, Bangkok. An international team led by general atomics (GA) is designing and constructing the new research complex. It comprises a 10 MW TRIGA type reactor, an isotope production and a centralized waste processing and storage facility. Electrowatt-Ekono Ltd. was hired by the Thai Government Agency, the Office of Atomic Energy for Peace (OAEP), as a consultant to the project. As the project is now approaching the end of its 4th year, it now stands at a decisive turning point. Basic design is nearly completed and detailed design is well advanced. The turnkey part of the contract including the reactor island, the isotope and waste facilities are still awaiting the issuance of the Construction Permit. Significant progress has been made on the other part of the project, which includes all the supporting infrastructure facilities. The Preliminary Safety Analysis Report (PSAR), prepared by GA, has been reviewed by various parties, including by nuclear safety experts from the IAEA, which has provided continuous support to the OAEP. Experts from the Argonne National Laboratory have been involved in the reviews as well. The PSAR is now under consideration at the Nuclear Facility Safety Sub-Committee (NFSS) of the Thai Atomic Energy for Peace Commission for issuing the Construction Permit of the ONRC Research Reactor. The following paper gives an overview of the project and its present status, outlining the features of the planned facilities and the issues the project is presently struggling with. Major lessons of the past 4 years are highlighted and an outlook into the future is attempted. (orig.)

  5. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  6. Research reactor decommissioning

    International Nuclear Information System (INIS)

    Full text: Of the ∼ 800 research reactors constructed worldwide to date, ∼50% have been shut down and are at various stages of decommissioning. Many reached the end of their design lives or were shut down due to strategic, economic or regulatory considerations. 27% of those in operation are over 40 years old and will need to be decommissioned soon. Decommissioning normally takes the facility permanently out of service and subjects it to progressive hazard reduction, dismantling and decontamination in a safe, secure economically viable way, using best practicable means to meet the best practicable environmental option, such that the risks and doses to workers and the general public are maintained as low as reasonably practicable. Whilst most decommissioning techniques are well established there are still some challenging and important issues that need resolution. Perhaps the most challenging issue is radioactive waste management and storage. It is vitally important that all local and national waste classification, transportation, storage and end point requirements are known, as the adopted strategy will be heavily influenced by these factors. Other equally important but softer issues include the requirement for early decommissioning plans, adequate funding/cost estimates and the involvement of all relevant stakeholders. A comprehensive decommissioning plan should be produced up front that encompasses an early radiological characterisation survey of the facility/site. An appropriate funding mechanism needs to be assured. Whilst regular revisions of the decommissioning cost study should help to determine required funds, it is important to validate these cost estimates by benchmarking other decommissioning projects and accumulated experience. The use of appropriate 'stakeholder dialogue' methods by the facility operator to inform and communicate with all interested parties, such as government and non-government organisations, regulators, trades unions, anti

  7. Research Reactors Coalitions

    International Nuclear Information System (INIS)

    When considering the potential role of an existing RR or possibly the construction of a new RR, it is clear that a nuclear science and technology programme (including nuclear power) could benefit provided the RR is safely and competently managed, well utilised and adequately funded. Based on MSs experience, a domestic RR may not be required to develop a nuclear power programme, provided the decision takes advantage of foreign expertise, including access to foreign RRs facilities and RRs regional/international networks. If a country decides to gain access to a foreign research reactor, it may need considering the potential risk of change in the political relationship with the host country that could compromise the achievement of its national relevant objectives. This risk may be offset by availability of many options within one or more regional/international RRs networks and coalitions. Examples include the use of existing RRs in vendor, non-vendor countries and, in some cases non-nuclear power countries, to develop human resources in support of the introduction of nuclear power elsewhere. International RR networking trends are most evident with high flux, higher capability, and more complex fuel and material testing RRs being shared through international partnerships. However, networks involving low-medium power RRs for education and training purposes are also gaining a more prominent role to support nuclear capacity building in newcomer MSs. Networking through the internet seems also to be a promising way to support, as complementary offer to direct access to RRs facilities, MSs nuclear capacity building objectives (e.g. the IRL project)

  8. Hiberarchy of requirement analysis of reactor protection system for advanced pressurized water reactor nuclear power plant

    International Nuclear Information System (INIS)

    In order to improve the security and the margin of safety of nuclear power plant, the research on requirement analysis of digital reactor protection system for advanced pressurized water reactor nuclear power plant was developed. Based on the known technology, a requirement analysis report was performed. A kind of three-levels pyramidal hierarchy was adopted in the requirement analysis, and the design characteristics of the requirement analysis were described in the analysis report. This hiberarchy can directly illuminate the design characters and logical achievement of the requirement analysis for advanced pressurized water reactor digital protection system. (authors)

  9. Meeting on reactor safety research

    International Nuclear Information System (INIS)

    The meeting 'Reactor Safety Research' organized for the second time by the GRS by order of the BMFT gave a review of research activities on the safety of light water reactors in the Federal Repulbic of Germany, international co-operation in this field and latest results of this research institution. The central fields of interest were subjects of man/machine-interaction, operational reliability accident sequences, and risk. (orig.)

  10. Conceptual design of the advanced marine reactor MRX

    International Nuclear Information System (INIS)

    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)

  11. Enrichment reduction for research reactors

    International Nuclear Information System (INIS)

    The worldwide activities on enrichment reduction for research reactors are reviewed and the national and international programs are described. Especially the following points are discussed: Benchmark calculations, reactor safety, fuel element development, irradiation tests, post irradiation examinations, full core demonstrations, activities of the GKSS and economical questions. (orig.)

  12. Ageing management for research reactors

    International Nuclear Information System (INIS)

    During the past several years, ageing of research reactor facilities continues to be an important safety issue. Despite the efforts exerted by operating organizations and regulatory authorities worldwide to address this issue, the need for an improved strategy as well as the need for establishing and implementing a systematic approach to ageing management at research reactors was identified. This paper discusses, on the basis of the IAEA Safety Standards, the effect of ageing on the safety of research reactors and presents a proactive strategy for ageing management. A systematic approach for ageing management is developed and presented together with its key elements, along with practical examples for their application. (author)

  13. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    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

  14. Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    GANTT, D.A.

    2000-01-12

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FETF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This revision reflects the 19 Oct 1999 baseline.

  15. Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    BOWEN, W.W.

    1999-11-08

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FFTF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This document reflects the 1 Oct 1999 baseline.

  16. IAEA safeguards at research reactors

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency applies safeguards to almost 150 facilities classified as research reactors. From a safeguards point of view, these facilities present a spectrum of features that must be addressed both from the nuclear material and from the operational viewpoints. The nuclear fuel used by these reactors varies from high enriched uranium (NEU), up to 93 U-235, to natural uranium and the thermal power output from over 100 megawatt to less than ten watts. Research reactors are also used for a wide variety of purposes, including materials testing, radiosotope production, training and nuclear physics studies. The effort spent by the Agency in safeguarding these reactors is dependant upon the thermal power of the reactor and on the quantity and type of nuclear material present. On some research reactors, the Agency devotes more inspection effort than on a large power reactor. On others, very little effort is reguired. Safeguards that are applied are done so according to Agency State agreements and consist of a combiination of nuclear material accounting and containment and surveillance. In this paper, the safeguards activities performed by the State and by the Agency will be reviewed for a large (≤50MWt) and for a small (≥ 1 MWt) reactor according to the most common type agreement. (author)

  17. Advanced reactor concepts and safety

    International Nuclear Information System (INIS)

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

  18. Development of inherent core technologies for advanced reactor

    International Nuclear Information System (INIS)

    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

  19. 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

  20. Light water reactor safety research

    International Nuclear Information System (INIS)

    As the technology of light water reactors (LWR) was being commercialized, the German Federal Government funded the reactor safety research program, which was conducted by national research centers, universities, and industry, and which led to the establishment, in early 1972, of the Nuclear Safety Project in Karlsruhe. In the seventies, the PNS project mainly studied the loss-of-coolant accident. Numerous experiments were run and computer codes developed for this purpose. In the eighties, the Karlsruhe Nuclear Research Center contributed to the German Risk Study, investigating especially core meltdown accidents under the impact of the events at Three Mile Island-2 and Chernobyl-4. Safety research in the nineties is concentrated on the requirements of future reactor generations, such as the European Pressurized Water Reactor (EPR) or potential approaches which, at the present time, are discernible only as tentative theoretical designs. (orig.)

  1. Advanced core monitoring technology for WWER reactors

    International Nuclear Information System (INIS)

    The Westinghouse BEACON online monitoring system has been developed to provide continuous core monitoring and operational support for pressurized water reactor using movable detectors (fission chamber) and core thermocouples. The basic BEACON core monitoring methodology is described. Traditional WWER reactors use rhodium fixed in-core detectors as the means to provide detailed core power distribution for surveillance purposes. An adapted version of the BEACON advanced core monitoring and support system is described which seems to be, due to the different demand/response requirements, the optimal solution (for routine surveillance and anomaly detection) for WWER reactors with existing fixed in-core detectors. (Z.S.) 4 refs

  2. Relative safeguards risks of advanced reactor concepts

    International Nuclear Information System (INIS)

    The purpose of this report is to develop a procedure to quantitatively assess the risk to diversion of the nuclear material in the fuel cycle for seven advanced reactor design concepts (LWR-Pu, LMFBR, HTGR, GCFR, MSBR, LWBR, HWR). Each stage in each of the seven reactor fuel cycles is evaluated and the result of the evaluation is a comparison among the various fuel cycles. This method of evaluation is used to determine the stages in a nuclear reactor fuel cycle that are most susceptible to diversion; it also gives an indication of what factors contribute to that susceptibility

  3. Advancement of light water reactor technology

    International Nuclear Information System (INIS)

    The Japanese technology of light water reactors is based on the technology imported from abroad around 1970, and the experience has been accumulated by the construction, operation and repair of light water reactors as well as the countermeasures to various troubles, moreover, the improvement and standardization of light water reactors have been promoted. As the result, recently the high capacity ratio has been attained, and the LWR technology has firmly taken root in Japan. The Subcommittee for the Advancement of Light Water Reactor Technology of the Advisory Committee for Energy has examined the subjects of technical development and the way the development should be in order to decide the strategy to advance LWR technology, and drawn up the interim report. The change of situation around the LWRs in Japan and the necessity to advance the technology, the target of advancing LWR technology and the subjects of the technical development, the system for the technical development and the securement of fund, and international cooperation are reported. The subjects of development are the pursuit of higher reliability and economic efficiency, the extension of plant life, the improvement of repairability and the reduction of radiation exposure, the improvement of operational capability, the reduction of wastes, the techniques for reactor decommissioning and the diversified location. (Kako, I.)

  4. The research reactors their contribution to the reactors physics

    International Nuclear Information System (INIS)

    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.)

  5. Facility modernization Annular Core Research Reactor

    International Nuclear Information System (INIS)

    The Annular Core Research Reactor (ACRR) has undergone numerous modifications since its conception in response to program needs. The original reactor fuel, which was special U-ZrH TRIGA fuel designed primarily for pulsing, has been replaced with a higher pulsing capacity BeO fuel. Other advanced operating modes which use this increased capability, in addition to the pulse and steady state, have been incorporated to tailor power histories and fluences to the experiments. Various experimental facilities have been developed that range from a radiography facility to a 50 cm diameter External Fuel Ring Cavity (FREC) using 180 of the original ZrH fuel elements. Currently a digital reactor console is being produced with GA, which will give enhanced monitoring capabilities of the reactor parameters while leaving the safety-related shutdown functions with analog technology. (author)

  6. Research and development long term prospects for advanced thermal and fast breeder reactors and fuel cycle activities

    International Nuclear Information System (INIS)

    Despite the present slowing down of most of the national nuclear programmes due to various conjunctural causes, nuclear energy is presumably to have in term an increasing role worldwide for electricity production, both in industrialized countries already equipped with nuclear power stations and in countries which intent to enter the nuclear energy route in a short or mid-term perspective. Nuclear energy development will have notably to take into account the following requirements: from a strategic point of view, the major concern is to optimize the use of world available natural uranium appears abundant and inexpensive, it must not be forgotten that in a longer time perspective a growing scarcity of this material will occur unavoidably; from the technical and economical standpoints, future reactors will have to be more reliable, more available, and even more safe, with lower KWh generating costs, than units presently constructed or operated. The paper presents the approach which is followed by France to meet these goals. (author)

  7. Determination of research reactor safety parameters by reactor calculations

    International Nuclear Information System (INIS)

    Main research reactor safety parameters such as power density peaking factors, shutdown margin and temperature reactivity coefficients are treated. Reactor physics explanation of the parameters is given together with their application in safety evaluation performed as part of research reactor operation. Reactor calculations are presented as a method for their determination assuming use of widely available computer codes. (author)

  8. Series lecture on advanced fusion reactors

    International Nuclear Information System (INIS)

    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.)

  9. Development of demonstration advanced thermal reactor

    International Nuclear Information System (INIS)

    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.)

  10. Experimental facility with two-phase flow and with high concentration of non-condensable gases for research and development of emergency cooling system of advanced nuclear reactors

    International Nuclear Information System (INIS)

    The development of emergency cooling passive systems of advanced nuclear reactors requires the research of some relative processes to natural circulation, in two-phase flow conditions involving condensation processes in the presence of non-condensable gases. This work describes the main characteristics of the experimental facility called Bancada de Circulacao Natural (BCN), designed for natural circulation experiments in a system with a hot source, electric heater, a cold source, heat exchanger, operating with two-phase flow and with high concentration of noncondensable gas, air. The operational tests, the data acquisition system and the first experimental results in natural circulation are presented. The experiments are transitory in natural circulation considering power steps. The distribution of temperatures and the behavior of the flow and of the pressure are analyzed. The experimental facility, the instrumentation and the data acquisition system demonstrated to be adapted for the purposes of research of emergency cooling passive systems, operating with two-phase flow and with high concentration of noncondensable gases. (author)

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

    International Nuclear Information System (INIS)

    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

  12. Development of the HANARO research reactor simulator for operator training

    International Nuclear Information System (INIS)

    HANARO (High flux Advanced Neutron Application ReactOr) is multi purpose research reactor in Korea Atomic Energy research Institute, and is operating since 1995. It is needed that training and retraining programs for the operating staff, including: reactor manager, shift supervisors, reactor operators, and others working at the research reactor facility. Recently, we developed HANARO research reactor real time simulator for operating staff training to satisfy these programs. The development of computer based training simulator have provided an easy understanding of reactor physics, operation, and control. Real time simulator is recognized as the ultimate training tool because they allow experiencing, in a dynamic mode, every type of operational condition which can be encountered including: start up, variation of power, shut down, operation during accidents, etc. Also, the simulator will be used as a dynamic test-bed for the reactor regulating system control algorithm

  13. Development of the HANARO research reactor simulator for operator training

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Kee Choon; Kim, Jang Yeol; Park, Jae Chang; Lee, Seung Wook; Hwang, In Ah; Lee, Dong Young [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    HANARO (High flux Advanced Neutron Application ReactOr) is multi purpose research reactor in Korea Atomic Energy research Institute, and is operating since 1995. It is needed that training and retraining programs for the operating staff, including: reactor manager, shift supervisors, reactor operators, and others working at the research reactor facility. Recently, we developed HANARO research reactor real time simulator for operating staff training to satisfy these programs. The development of computer based training simulator have provided an easy understanding of reactor physics, operation, and control. Real time simulator is recognized as the ultimate training tool because they allow experiencing, in a dynamic mode, every type of operational condition which can be encountered including: start up, variation of power, shut down, operation during accidents, etc. Also, the simulator will be used as a dynamic test-bed for the reactor regulating system control algorithm.

  14. Development of Korea advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Future nuclear power plants should not only have the features of improved safety and economic competitiveness but also provide a means to resolve spent fuel storage problems by minimizing volume of high level wastes. It is widely believed that liquid metal reactors (LMRs) have the highest potential of meeting these requirements. In this context, the LMR development program was launched as a national long-term R and D program in 1992, with a target to introduce a commercial LMR around 2030. Korea Advanced Liquid Metal Reactor (KALIMER), a 150 MWe pool-type sodium cooled prototype reactor, is currently under the conceptual design study with the target schedule to complete its construction by the mid-2010s. This paper summarizes the KALIMER development program and major technical features of the reactor system. (author)

  15. Advanced light water reactor plant

    International Nuclear Information System (INIS)

    For nuclear power to be competitive with the other methods of electrical power generation the economic performance should be significantly improved by increasing the time spent on line generating electricity relative to time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described which is used to resolve maintenance related operating cycle length barriers. Advanced light water nuclear power plant is designed with the purpose to maximize online generating time by increasing operating cycle length. (author)

  16. The utilisation of thorium fuel in a generation 4 light water reactor design

    International Nuclear Information System (INIS)

    During the last several years the Department of Energy has sponsored research at Purdue University on advanced reactor designs under the Nuclear Energy Research Initiative (NERI) programme. This work has involved research in 'Generation IV' advanced reactor designs such as a high conversion boiling water reactor, as well research in advanced fuel designs such a metal matrix 'dispersion' fuel. The unifying theme of this research has been to take advantage of the numerous benefits of the thorium fuel cycle. The Purdue research has been performed in collaboration with Argonne and Brookhaven National Laboratories for the dispersion fuel research and the high conversion reactor research, respectively. The primary contribution to both research efforts from Purdue has been on neutronics design and analysis. This paper will focus on the neutronics design and analysis of the high conversion boiling water reactor. (author)

  17. Advanced fuel cycles for CANDU reactors

    International Nuclear Information System (INIS)

    The current natural uranium-fuelled CANDU system is a world leader, both in terms of overall performance and uranium utilization. Moreover, the CANDU reactor is capable of using many different advanced fuel cycles, with improved uranium utilization relative to the natural uranium one-through cycle. This versatility would enable CANDU to maintain its competitive edge in uranium utilization as improvements are made by the competition. Several CANDU fuel cycles are symbiotic with LWRs, providing an economical vehicle for the recycle of uranium and/or plutonium from discharges LWR fuel. The slightly enriched uranium (SEU) fuel cycle is economically attractive now, and this economic benefit will increase with anticipated increases in the cost of natural uranium, and decreases in the cost of fuel enrichment. The CANFLEX fuel bundle, an advanced 43-element design, will ensure that the full benefits of SEU, and other advanced fuel cycles, can be achieved in the CANDU reactor. 25 refs

  18. The neutron scattering platform of China's Advanced Research Reactor%中国先进研究堆中子散射科学平台介绍*

    Institute of Scientific and Technical Information of China (English)

    刘蕴韬; 陈东风

    2013-01-01

    As a very important method to discover the world neutron scattering technique has been applied in many fields and achieved a lot. Based on the neutron scattering facilities just built at ChinaAdvanced Research Reactor the principle and characters of this technique and facilities are intro-duced in details, and their prospected applications in the future are also described.%  中子散射技术作为人类认知世界不可或缺的独特手段,多年来在诸多领域得到了广泛应用并成绩显著。文章以新建成的中国先进研究堆中子散射科学平台谱仪为例,较为详细地介绍了中子散射技术和谱仪的基本原理和特点,并对其未来的应用进行了展望。

  19. Research reactor education and training

    International Nuclear Information System (INIS)

    CORYS T.E.S.S. and TECHNICATOME present in this document some of the questions that can be rightfully raised concerning education and training of nuclear facilities' staffs. At first, some answers illustrate the tackled generic topics: importance of training, building of a training program, usable tools for training purposes. Afterwards, this paper deals more specifically with research reactors as an actual training tool. The pedagogical advantages they can bring are illustrated through an example consisting in the description of the AZUR facility training capabilities followed by the detailed experiences CORYS T.E.S.S. and TECHNICATOME have both gathered and keeps on gaining using research reactors for training means. The experience shows that this incomparable training material is not necessarily reserved to huge companies or organisations' numerous personnel. It offers enough flexibility to be adapted to the specific needs of a thinner audience. Thus research reactor staffs can also take advantages of this training method. (author)

  20. Engineering design of advanced marine reactor MRX

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-01

    JAERI has studied the design of an advanced marine reactor (named as MRX), which meets requirements of the enhancement of economy and reliability, by reflecting results and knowledge obtained from the development of N.S. Mutsu. The MRX with a power of 100 MWt is intended to be used for ship propulsion such as an ice-breaker, container cargo ship and so on. After completion of the conceptual design, the engineering design was performed in four year plan from FY 1993 to 1996. (1) Compactness, light-weightiness and simplicity of the reactor system are realized by adopting an integral-type PWR, i.e. by installing the steam generator, the pressurizer, and the control rod drive mechanism (CRDM) inside the pressure vessel. Because of elimination of the primary coolant circulation pipes in the MRX, possibility of large-scale pipe break accidents can be eliminated. This contributes to improve the safety of the reactor system and to simplify the engineered safety systems. (2) The in-vessel type CRDM contributes not only to eliminate possibilities of rod ejection accidents, but also to make the reactor system compact. (3) The concept of water-filled containment where the reactor pressure vessel is immersed in the water is adopted. It can be of use for emergency core cooling system which maintains core flooding passively in case of a loss-of-coolant accident. The water-filled containment system also contributes essentially light-weightness of the reactor system since the water inside containment acts as a radiation shield and in consequence the secondary radiation shield can be eliminated. (4) Adoption of passive decay heat removal systems has contributed in a greater deal to simplification of the engineered safety systems and to enhancement of reliability of the systems. (5) Operability has been improved by simplification of the whole reactor system, by adoption of the passive safety systems, advanced automatic operation systems, and so on. (J.P.N.)

  1. Local AREA networks in advanced nuclear reactors

    International Nuclear Information System (INIS)

    The report assesses Local Area Network Communications with a view to their application in advanced nuclear reactor control and protection systems. Attention is focussed on commercially available techniques and systems for achieving the high reliability and availability required. A basis for evaluating network characteristics in terms of broadband or baseband type, medium, topology, node structure and access method is established. The reliability and availability of networks is then discussed. Several commercial networks are briefly assessed and a distinction made between general purpose networks and those suitable for process control. The communications requirements of nuclear reactor control and protection systems are compared with the facilities provided by current technology

  2. Uncertainty quantification approaches for advanced reactor analyses.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Nuclear Engineering Division

    2009-03-24

    The original approach to nuclear reactor design or safety analyses was to make very conservative modeling assumptions so as to ensure meeting the required safety margins. Traditional regulation, as established by the U. S. Nuclear Regulatory Commission required conservatisms which have subsequently been shown to be excessive. The commission has therefore moved away from excessively conservative evaluations and has determined best-estimate calculations to be an acceptable alternative to conservative models, provided the best-estimate results are accompanied by an uncertainty evaluation which can demonstrate that, when a set of analysis cases which statistically account for uncertainties of all types are generated, there is a 95% probability that at least 95% of the cases meet the safety margins. To date, nearly all published work addressing uncertainty evaluations of nuclear power plant calculations has focused on light water reactors and on large-break loss-of-coolant accident (LBLOCA) analyses. However, there is nothing in the uncertainty evaluation methodologies that is limited to a specific type of reactor or to specific types of plant scenarios. These same methodologies can be equally well applied to analyses for high-temperature gas-cooled reactors and to liquid metal reactors, and they can be applied to steady-state calculations, operational transients, or severe accident scenarios. This report reviews and compares both statistical and deterministic uncertainty evaluation approaches. Recommendations are given for selection of an uncertainty methodology and for considerations to be factored into the process of evaluating uncertainties for advanced reactor best-estimate analyses.

  3. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Baldev Raj

    2009-06-01

    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 of providing fast reactor electricity at an affordable and competitive price.

  4. Advanced Fast Reactor - 100 - Design Overview

    International Nuclear Information System (INIS)

    The Advanced Fast Reactor-100 (AFR-100) is a small modular sodium-cooled fast reactor with an electrical power output of 100MWe. The AFR-100 has a long-lived core that does not require refueling for 30 years. The concept contains various innovations such as a small compact modular core (both vented and non-vented fuel pins), advanced core shielding materials, a compact fuel handling system, advanced electromagnetic pumps, compact intermediate heat exchangers, and a direct reactor auxiliary cooling system. These advanced systems and components were adopted in order to reduce the overall size of the primary heat transport system (and therefore the overall commodities and cost), enhance safety, and to improve overall plant performance. This paper presents the summary results of a year-long study that culminated in the design of two primary heat transport configurations for the AFR-100. The paper describes those innovations and shows how they are integrated into the overall AFR-100 primary heat transport system design. (author)

  5. Research reactor modernization and refurbishment

    International Nuclear Information System (INIS)

    Many recent, high profile research reactor unplanned shutdowns can be directly linked to different challenges which have evolved over time. The concept of ageing management is certainly nothing new to nuclear facilities, however, these events are highlighting the direct impact unplanned shutdowns at research reactors have on various stakeholders who depend on research reactor goods and services. Provided the demand for these goods and services remains strong, large capital projects are anticipated to continue in order to sustain future operation of many research reactors. It is within this context that the IAEA organized a Technical Workshop to launch a broader Agency activity on research reactor modernization and refurbishment (M and R). The workshop was hosted by the operating organization of the HOR Research Reactor in Delft, the Netherlands, in October 2006. Forty participants from twenty-three countries participated in the meeting: with representation from Africa, Asia Pacific, Eastern Europe, North America, South America and Western Europe. The specific objectives of this workshop were to present facility reports on completed, existing and planned M and R projects, including the project objectives, scope and main characteristics; and to specifically report on: - the project impact (planned or actual) on the primary and key supporting motivation for the M and R project; - the project impact (planned or actual) on the design basis, safety, and/or regulatory-related reports; - the project impact (planned or actual) on facility utilization; - significant lessons learned during or following the completion of M and R work. Contributions from this workshop were reviewed by experts during a consultancy meeting held in Vienna in December 2007. The experts selected final contributions for inclusion in this report. Requests were also distributed to some authors for additional detail as well as new authors for known projects not submitted during the initial 2006 workshop

  6. Thermochemical modelling of advanced CANDU reactor fuel

    Science.gov (United States)

    Corcoran, Emily Catherine

    2009-04-01

    With an aging fleet of nuclear generating facilities, the imperative to limit the use of non-renewal fossil fuels and the inevitable need for additional electricity to power Canada's economy, a renaissance in the use of nuclear technology in Canada is at hand. The experience and knowledge of over 40 years of CANDU research, development and operation in Ontario and elsewhere has been applied to a new generation of CANDU, the Advanced CANDU Reactor (ACR). Improved fuel design allows for an extended burnup, which is a significant improvement, enhancing the safety and the economies of the ACR. The use of a Burnable Neutron Absorber (BNA) material and Low Enriched Uranium (LEU) fuel has created a need to understand better these novel materials and fuel types. This thesis documents a work to advance the scientific and technological knowledge of the ACR fuel design with respect to thermodynamic phase stability and fuel oxidation modelling. For the BNA material, a new (BNA) model is created based on the fundamental first principles of Gibbs energy minimization applied to material phase stability. For LEU fuel, the methodology used for the BNA model is applied to the oxidation of irradiated fuel. The pertinent knowledge base for uranium, oxygen and the major fission products is reviewed, updated and integrated to create a model that is applicable to current and future CANDU fuel designs. As part of this thesis, X-Ray Diffraction (XRD) and Coulombic Titration (CT) experiments are compared to the BNA and LEU models, respectively. From the analysis of the CT results, a number of improvements are proposed to enhance the LEU model and provide confidence in its application to ACR fuel. A number of applications for the potential use of these models are proposed and discussed. Keywords: CANDU Fuel, Gibbs Energy Mimimization, Low Enriched Uranium (LEU) Fuel, Burnable Neutron Absorber (BNA) Material, Coulometric Titration, X-Ray Diffraction

  7. Upgrading the NRU research reactor

    International Nuclear Information System (INIS)

    After a nearly two-year long detailed review, AECL Research decided that its NRU research reactor will complete its mission around the turn of the century. The company's original intentions for major refurbishment have been revised and upgrading work will now mainly comprise add-ons to existing systems - so that research projects and isotope production schedules can be met - and procedure modifications to ensure continued safe operation. (Author)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  10. Mirror Advanced Reactor Study interim design report

    International Nuclear Information System (INIS)

    The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design

  11. Mirror Advanced Reactor Study interim design report

    Energy Technology Data Exchange (ETDEWEB)

    1983-04-01

    The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design.

  12. Research reactors: design, safety requirements and applications

    International Nuclear Information System (INIS)

    There are two types of reactors: research reactors or power reactors. The difference between the research reactor and energy reactor is that the research reactor has working temperature and fuel less than the power reactor. The research reactors cooling uses light or heavy water and also research reactors need reflector of graphite or beryllium to reduce the loss of neutrons from the reactor core. Research reactors are used for research training as well as testing of materials and the production of radioisotopes for medical uses and for industrial application. The difference is also that the research reactor smaller in terms of capacity than that of power plant. Research reactors produce radioactive isotopes are not used for energy production, the power plant generates electrical energy. In the world there are more than 284 reactor research in 56 countries, operates as source of neutron for scientific research. Among the incidents related to nuclear reactors leak radiation partial reactor which took place in three mile island nuclear near pennsylvania in 1979, due to result of the loss of control of the fission reaction, which led to the explosion emitting hug amounts of radiation. However, there was control of radiation inside the building, and so no occurred then, another accident that lead to radiation leakage similar in nuclear power plant Chernobyl in Russia in 1986, has led to deaths of 4000 people and exposing hundreds of thousands to radiation, and can continue to be effect of harmful radiation to affect future generations. (author)

  13. Development of advanced nuclear reactors in Russia

    International Nuclear Information System (INIS)

    Several advanced reactor designs have been so far developed in Russia. The AES-91 and AES-92 plants with the VVER-1000 reactors have been developed at the beginning of 1990. However, the former design has been built in China and the latest which is certified meeting European Utility Requirements is being built in India. Moreover, the model VVER-1500 reactor with 50-60 MWd/t burn-up and an enhanced safety was being developed by Gidropress about 2005, excepting to be completed in 2007. But, this schedule has slipped in favor of development of the AES-2006 power plant incorporating a third-generation standardized VVER-1200 reactor of 1170 MWe. This is an evolutionary development of the well-proven VVER-1000 reactor in the AES-92 plant, with longer life, greater power and efficiency and its lead units are being built at Novovoronezh II, to start operation in 2012-13. Based on Atomenergoproekt declaration, the AES-2006 conforms to both Russian standards and European Utility Requirements. The most important features of the AES-2006 design are mentioned as: a design based on the passive safety systems, double containment, longer plant service life of 50 years with a capacity factor of 92%, longer irreplaceable components service life of 60 years, a 28.6% lower amount of concrete and metal, shorter construction time of 54 months, a Core Damage Frequency of 1x10-7/ year and lower liquid and solid wastes by 70% and 80% respectively. The presented paper includes a comparative analysis of technological and safety features, economic parameters and environmental impact of the AES-2006 design versus the other western advanced reactors. Since the Bushehr phase II NPP and several other NPPs are planning in Iran, such analysis would be of a great importance

  14. Advanced nuclear reactor systems - an Indian perspective

    International Nuclear Information System (INIS)

    The Indian nuclear power programme envisages use of closed nuclear fuel cycle and thorium utilisation as its mainstay for its sustainable growth. The current levels of deployment of nuclear energy in India need to be multiplied nearly hundred fold to reach levels of electricity generation that would facilitate the country to achieve energy independence as well as a developed status. The Indian thorium based nuclear energy systems are being developed to achieve sustainability in respect of fuel resource along with enhanced safety and reduced waste generation. Advanced Heavy Water Reactor and its variants have been designed to meet these objectives. The Indian High Temperature Reactor programme also envisages use of thorium-based fuel with advanced levels of passive safety features. (author)

  15. Advanced reactor physics methods for heterogeneous reactor cores

    Science.gov (United States)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  16. Fundamental Thermal Fluid Physics of High Temperature Flows in Advanced Reactor Systems - Nuclear Energy Research Initiative Program Interoffice Work Order (IWO) MSF99-0254 Final Report for Period 1 August 1999 to 31 December 2002

    Energy Technology Data Exchange (ETDEWEB)

    McEligot, D.M.; Condie, K.G.; Foust, T.D.; McCreery, G.E.; Pink, R.J.; Stacey, D.E. (INEEL); Shenoy, A.; Baccaglini, G. (General Atomics); Pletcher, R.H. (Iowa State U.); Wallace, J.M.; Vukoslavcevic, P. (U. Maryland); Jackson, J.D. (U. Manchester, UK); Kunugi, T. (Kyoto U., Japan); Satake, S.-i. (Tokyo U. Science, Japan)

    2002-12-31

    The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of advanced reactors for higher efficiency and enhanced safety and for deployable reactors for electrical power generation, process heat utilization and hydrogen generation. While key applications would be advanced gas-cooled reactors (AGCRs) using the closed Brayton cycle (CBC) for higher efficiency (such as the proposed Gas Turbine - Modular Helium Reactor (GT-MHR) of General Atomics [Neylan and Simon, 1996]), results of the proposed research should also be valuable in reactor systems with supercritical flow or superheated vapors, e.g., steam. Higher efficiency leads to lower cost/kwh and reduces life-cycle impacts of radioactive waste (by reducing waters/kwh). The outcome will also be useful for some space power and propulsion concepts and for some fusion reactor concepts as side benefits, but they are not the thrusts of the investigation. The objective of the project is to provide fundamental thermal fluid physics knowledge and measurements necessary for the development of the improved methods for the applications.

  17. Advanced Neutron Source Reactor thermal analysis of fuel plate defects

    International Nuclear Information System (INIS)

    The Advanced Neutron Source Reactor (ANSR) is a research reactor designed to provide the highest continuous neutron beam intensity of any reactor in the world. The present technology for determining safe operations were developed for the High Flux Isotope Reactor (HFIR). These techniques are conservative and provide confidence in the safe operation of HFIR. However, the more intense requirements of ANSR necessitate the development of more accurate, but still conservative, techniques. This report details the development of a Local Analysis Technique (LAT) that provides an appropriate approach. Application of the LAT to two ANSR core designs are presented. New theories of the thermal and nuclear behavior of the U3Si2 fuel are utilized. The implications of lower fuel enrichment and of modifying the inspection procedures are also discussed. Development of the computer codes that enable the automate execution of the LAT is included

  18. Australia's replacement research reactor project

    International Nuclear Information System (INIS)

    HIFAR, a 10 MW tank type DIDO Class reactor has operated at the Lucas Heights Science and Technology Centre for 43 years. HIFAR and the 10 kW Argonaut reactor 'Moata' which is in the Care and Maintenance phase of decommissioning are Australia's only nuclear reactors. The initial purpose for HIFAR was for materials testing to support a nuclear power program. Changing community attitude through the 1970's and a Government decision not to proceed with a planned nuclear power reactor resulted in a reduction of materials testing activities and a greater emphasis being placed on neutron beam research and the production of radioisotopes, particularly for medical purposes. HIFAR is not fully capable of satisfying the expected increase in demand for medical radiopharmaceuticals beyond the next 5 years and the radial configuration of the beam tubes severely restricts the scope and efficiency of neutron beam research. In 1997 the Australian Government decided that a replacement research reactor should be built by the Australian Nuclear Science and Technology Organisation at Lucas Heights subject to favourable results of an Environmental Impact Study. The Ei identified no reasons on the grounds of safety, health, hazard or risk to prevent construction on the preferred site and it was decided in May 1999 that there were no environmental reasons why construction of the facility should not proceed. In recent years ANSTO has been reviewing the operation of HIFAR and observing international developments in reactor technology. Limitations in the flexibility and efficiency achievable in operation of a tank type reactor and the higher intrinsic safety sought in fundamental design resulted in an early decision that the replacement reactor must be a pool type having cleaner and higher intensity tangential neutron beams of wider energy range than those available from HIFAR. ANSTO has chosen to use it's own resources supported by specialised external knowledge and experience to identify

  19. Advanced gas cooled reactors - Designing for safety

    International Nuclear Information System (INIS)

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme

  20. 2012 review of French research reactors

    International Nuclear Information System (INIS)

    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)

  1. Light-Water-Reactor Safety Research Program. Quarterly progress report, January--March 1978

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-05-01

    This progress report summarizes the Argonne National Laboratory work performed during January, February, and March 1978 on water-reactor-safety problems. The following research and development areas are covered: (1) Loss-of-coolant Accident Research: Heat Transfer and Fluid Dynamics; (2) Transient Fuel Response and Fission-product Release Program; (3) Mechanical Properties of Zircaloy Containing Oxygen; and (4) Steam-explosion Studies.

  2. Fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    This paper includes some statements and remarks concerning the uranium silicide fuels for which there is significant fabrication in AECL, irradiation and defect performance experience; description of two Canadian high flux research reactors which use high enrichment uranium (HEU) and the fuels currently used in these reactors; limited fabrication work done on Al-U alloys to uranium contents as high as 40 wt%. The latter concerns work aimed at AECL fast neutron program. This experience in general terms is applied to the NRX and NRU designs of fuel

  3. Materials research with neutron beams from a research reactor

    International Nuclear Information System (INIS)

    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)

  4. 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)

  5. Advanced Recycling Reactor with Minor Actinide Fuel

    International Nuclear Information System (INIS)

    The Advanced Recycling Reactor (ARR) with minor actinide fuel has been studied. This paper presents the pre-conceptual design of the ARR proposed by the International Nuclear Recycling Alliance (INRA) for FOA study sponsored by DOE of the United States of America (U.S.). Although the basic reactor concept is technically mature, it is not suitable for commercial use due to the need to reduce capital costs. As a result of INRA's extensive experience, it is anticipated that a non-commercial ARR1 will be viable and meet U.S. requirements by 2025. Commercial Advanced Recycling Reactor (ARR) operations are expected to be feasible in competition with LWRs by 2050, based on construction of ARR2 in 2035. The ARR based on the Japan Sodium-cooled Fast Reactor (JSFR) is a loop-typed sodium cooled reactor with MOX fuel that is selected because of much experience of SFRs in the world. Major features of key technology enhancements incorporated into the ARR are the following: Decay heat can be removed by natural circulation to improve safety. The primary cooling system consists of two-loop system and the integrated IHX/Pump to improve economics. The steam generator with the straight double-walled tube is used to improve reliability. The reactor core of the ARR1 is 70 cm high and the volume fraction of fuel is 31.6%. The conversion ratio of fissile is set up less than 0.65 and the amount of burned TRU is 45-51 kg/TWeh. According to survey of more effective TRU burning core, the oxide fuel core containing high TRU (MA 15%, Pu 35% average) with moderate pins of 12% arranged driver fuel assemblies can decrease TRU conversion ratio to 0.33 and improve TRU burning capability to 67 kg/TWeh. The moderator can enhance TRU burning, while increasing the Doppler effect and reducing the positive sodium void effect. High TRU fraction promotes TRU burning by curbing plutonium production. High Am fraction and Am blanket promote Am transmutation. The ARR1 consists of a reactor building (including

  6. Present status of research reactor decommissioning programme in Indonesia

    International Nuclear Information System (INIS)

    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)

  7. Fast reactors and advanced light water reactors for sustainable development

    International Nuclear Information System (INIS)

    Complete text of publication follows: The importance of nuclear energy, as a realistic option to solve the issues of the depletion of energy resources and the global environment, has been re-acknowledged worldwide. In response to this international movement, the papers compiling the most recent findings in the fields of fast reactors (FR) and advanced light water reactors (LWR) were gathered and published in this special issue. This special issue compiles six articles, most of which are very meticulously performed studies of the multi year development of design and assessment methods for large sodium-cooled FRs (SFRs), and two are related to the fuel cycle options that are leading to a greater understanding on the efficient utilization of energy resources. The Japanese sodium-cooled fast reactor (JSFR) is addressed in two manuscripts. H. Yamano et al. reviewed the current design which adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. Their safety assessments of both design basis accidents and severe accidents indicate that the devised JSFR satisfies well their risk target. T. Takeda et al. discussed the improvement of the modeling accuracy for the detailed calculation of JSFR's features in three areas: neutronics, fuel materials, and thermal hydraulics. The verification studies which partly use the measured data from the prototype FBR Monju are also described. Two of these manuscripts deal with those aspects of advanced design of SFR that have hitherto not been explored in great depth. The paper by G. Palmiotti et al. explored the possibility of using the sensitivity methodologies in the reactor physics field. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described. F. Baque et al. reviewed the evolution of the in

  8. Johnson Noise Thermometry for Advanced Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Britton, C.L.,Jr.; Roberts, M.; Bull, N.D.; Holcomb, D.E.; Wood, R.T.

    2012-09-15

    Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor’s physical condition. In and near the core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of the current ORNL-led project, conducted under the Instrumentation, Controls, and Human-Machine Interface (ICHMI) research pathway of the U.S. Department of Energy (DOE) Advanced SMR Research and Development (R&D) program, is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

  9. The korea multi-purpose research reactor

    International Nuclear Information System (INIS)

    This paper presents and discusses background and status of the design of the 30MW Korea Multi-purpose Research Reactor(KMRR) which is planed to achieve its first criticality in December, 19992, at Daeduk site of the Korea Advanced Energy Research Institute (KAERI). KAERI playing the leading role in Korea's nuclear technology development takes the total responsibility for its design, construction and operation. Number of Korean nuclear industries are, also, actively participating in the project while making the most of their expertise in relevant areas. (Author)

  10. ACR-1000TM - advanced Candu reactor design

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has developed the Advanced CANDU ReactorTM- 1000 (ACR-1000TM) as an evolutionary advancement of the current CANDU 6TM reactor. This evolutionary advancement is based on AECL's in-depth knowledge of CANDU structures, systems, components and materials, gained during 50 years of continuous construction, engineering and commissioning, as well as on the experience and feedback received from operators of CANDU plants. The ACR design retains the proven strengths and features of CANDU reactors, while incorporating innovations and state-of-the-art technology. These innovations improve economics, inherent safety characteristics, and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. The Canadian nuclear reactor design evolution that has reached today's stage represented by the ACR-1000, has a long history dating back to the early 1950's. In this regard, Canada is in a unique situation, shared only by a very few other countries, where original nuclear power technology has been invented and further developed. The ACR design has been reviewed by domestic and international regulatory bodies, and has been given a positive regulatory opinion about its licensability. The Canadian regulator, the Canadian Nuclear Safety Commission (CNSC) completed the Phase 1 and Phase 2 pre-project design reviews in December 2008 and August 2009, respectively, and concluded that there are no fundamental barriers to licensing the ACR-1000 design in Canada. The final stage of the ACR-1000 design is currently underway and will be completed by fall of 2011, along with the final elements of the safety analyses and probabilistic safety analyses supporting the finalized design. The generic Preliminary Safety Analysis Report (PSAR) for the ACR-1000 was completed in September 2009. The PSAR demonstrates ACR-1000 safety case and compliance with Canadian and international regulatory requirements and expectations. (authors)

  11. Experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source spectrum of the NBSR reactor at the NIST Center for Neutron Research

    Energy Technology Data Exchange (ETDEWEB)

    Cook, J.C.; Barker, J.G.; Rowe, J.M.; Williams, R.E. [NIST Center for Neutron Research, National Institute of Standards and Technology, 100 Bureau Drive, Mail Stop 6100, Gaithersburg, MD 20899-6100 (United States); Gagnon, C. [Department of Materials Science and Engineering, University of Maryland, College Park, MD 20742 (United States); Lindstrom, R.M. [Scientist Emeritus, Chemical Sciences Division, National Institute of Standards and Technology, 100 Bureau Drive, Mail Stop 8395, Gaithersburg, MD 20899-8395 (United States); Ibberson, R.M.; Neumann, D.A. [NIST Center for Neutron Research, National Institute of Standards and Technology, 100 Bureau Drive, Mail Stop 6100, Gaithersburg, MD 20899-6100 (United States)

    2015-08-21

    The recent expansion of the National Institute of Standards and Technology (NIST) Center for Neutron Research facility has offered a rare opportunity to perform an accurate measurement of the cold neutron spectrum at the exit of a newly-installed neutron guide. Using a combination of a neutron time-of-flight measurement, a gold foil activation measurement, and Monte Carlo simulation of the neutron guide transmission, we obtain the most reliable experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source brightness to date. Time-of-flight measurements were performed at three distinct fuel burnup intervals, including one immediately following reactor startup. Prior to the latter measurement, the hydrogen was maintained in a liquefied state for an extended period in an attempt to observe an initial radiation-induced increase of the ortho (o)-hydrogen fraction. Since para (p)-hydrogen has a small scattering cross-section for neutron energies below 15 meV (neutron wavelengths greater than about 2.3 Å), changes in the o- p hydrogen ratio and in the void distribution in the boiling hydrogen influence the spectral distribution. The nature of such changes is simulated with a continuous-energy, Monte Carlo radiation-transport code using 20 K o and p hydrogen scattering kernels and an estimated hydrogen density distribution derived from an analysis of localized heat loads. A comparison of the transport calculations with the mean brightness function resulting from the three measurements suggests an overall o- p ratio of about 17.5(±1) % o- 82.5% p for neutron energies<15 meV, a significantly lower ortho concentration than previously assumed.

  12. Experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source spectrum of the NBSR reactor at the NIST Center for Neutron Research

    Science.gov (United States)

    Cook, J. C.; Barker, J. G.; Rowe, J. M.; Williams, R. E.; Gagnon, C.; Lindstrom, R. M.; Ibberson, R. M.; Neumann, D. A.

    2015-08-01

    The recent expansion of the National Institute of Standards and Technology (NIST) Center for Neutron Research facility has offered a rare opportunity to perform an accurate measurement of the cold neutron spectrum at the exit of a newly-installed neutron guide. Using a combination of a neutron time-of-flight measurement, a gold foil activation measurement, and Monte Carlo simulation of the neutron guide transmission, we obtain the most reliable experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source brightness to date. Time-of-flight measurements were performed at three distinct fuel burnup intervals, including one immediately following reactor startup. Prior to the latter measurement, the hydrogen was maintained in a liquefied state for an extended period in an attempt to observe an initial radiation-induced increase of the ortho (o)-hydrogen fraction. Since para (p)-hydrogen has a small scattering cross-section for neutron energies below 15 meV (neutron wavelengths greater than about 2.3 Å), changes in the o- p hydrogen ratio and in the void distribution in the boiling hydrogen influence the spectral distribution. The nature of such changes is simulated with a continuous-energy, Monte Carlo radiation-transport code using 20 K o and p hydrogen scattering kernels and an estimated hydrogen density distribution derived from an analysis of localized heat loads. A comparison of the transport calculations with the mean brightness function resulting from the three measurements suggests an overall o- p ratio of about 17.5(±1) % o- 82.5% p for neutron energies<15 meV, a significantly lower ortho concentration than previously assumed.

  13. 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

  14. Research reactor's role in Korea

    International Nuclear Information System (INIS)

    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

  15. New research reactor for Australia

    International Nuclear Information System (INIS)

    HIFAR, Australia's major research reactor was commissioned in 1958 to test materials for an envisaged indigenous nuclear power industry. HIFAR is a Dido type reactor which is operated at 10 MW. With the decision in the early 1970's not to proceed to nuclear power, HIFAR was adapted to other uses and has served Australia well as a base for national nuclear competence; as a national facility for neutron scattering/beam research; as a source of radioisotopes for medical diagnosis and treatment; and as a source of export revenue from the neutron transmutation doping of silicon for the semiconductor industry. However, all of HIFAR's capabilities are becoming less than optimum by world and regional standards. Neutron beam facilities have been overtaken on the world scene by research reactors with increased neutron fluxes, cold sources, and improved beams and neutron guides. Radioisotope production capabilities, while adequate to meet Australia's needs, cannot be easily expanded to tap the growing world market in radiopharmaceuticals. Similarly, neutron transmutation doped silicon production, and export income from it, is limited at a time when the world market for this material is expanding. ANSTO has therefore embarked on a program to replace HIFAR with a new multi-purpose national facility for nuclear research and technology in the form of a reactor: a) for neutron beam research, - with a peak thermal flux of the order of three times higher than that from HIFAR, - with a cold neutron source, guides and beam hall, b) that has radioisotope production facilities that are as good as, or better than, those in HIFAR, c) that maximizes the potential for commercial irradiations to offset facility operating costs, d) that maximizes flexibility to accommodate variations in user requirements during the life of the facility. ANSTO's case for the new research reactor received significant support earlier this month with the tabling in Parliament of a report by the Australian Science

  16. The research reactor TRIGA Mainz

    International Nuclear Information System (INIS)

    The TRIGA Mark II reactor at the Institut fuer Kernchemie became first critical on August 3rd, 1965. It can be operated in the steady state mode with a maximum power of 100 kWth and in the pulse mode with a peak power of 250 MWth. A survey of the research programmes performed at the TRIGA Mainz is given covering applications in basic research as well as applied science in nuclear chemistry and nuclear physics. Furthermore, the reactor is used for neutron activation analysis and for education and training of scientists, teachers, students and technical personal. Important projects for the future of the TRIGA Mainz are the UCN (ultra cold neutrons) experiment, fast chemical separation, medical applications and the use of the NAA as well as the use of the reactor facility for the training of students in the fields of nuclear chemistry, nuclear physics and radiation protection. Taking into account the past and future operation schedule and the typically low burn-up of TRIGA fuel elements (∝4 g U-235/a), the reactor can be operated for at least the next decade taking into account the fresh fuel elements on stock and without changing spent fuels. (orig.)

  17. 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.

  18. Advanced integral reactor (SMART) for nuclear desalination

    International Nuclear Information System (INIS)

    At present, severe fresh water shortages are occurring in some regional areas of the Republic of Korea and the problem is expected to spread throughout the country within a decade unless appropriate and timely countermeasures are taken. Of these, nuclear sea water desalination is receiving much attention because the Republic of Korea has a firmly established nuclear environment and abundant sea water resources. In addition, nuclear plants provide cleaner energy than fossil plants, which is another important beneficial factor for countries as crowded as ours. With a view to applying nuclear desalination, development of SMART (system integrated modular advanced reactor) was initiated and is currently in progress. SMART is being developed as a 330 MW(th) integral reactor with passive safety features. The design of SMART is aimed at combining the firmly established commercial reactor design with new advanced technologies. This has led to the use of industry proven Korea optimized fuel assembly (KOFA) based fuels, while radically new technologies such as a self-pressurizing pressurizer, helical once-through steam generators and a new control concept are being developed. The current development status of SMART and its application to nuclear desalination are presented. (author)

  19. Next generation advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    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.

  20. Surveys of research in the Chemistry Division, Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Grazis, B.M. [ed.

    1992-11-01

    Research reports are presented on reactive intermediates in condensed phase (radiation chemistry, photochemistry), electron transfer and energy conversion, photosynthesis and solar energy conversion, metal cluster chemistry, chemical dynamics in gas phase, photoionization-photoelectrons, characterization and reactivity of coal and coal macerals, premium coal sample program, chemical separations, heavy elements coordination chemistry, heavy elements photophysics/photochemistry, f-electron interactions, radiation chemistry of high-level wastes (gas generation in waste tanks), ultrafast molecular electronic devices, and nuclear medicine. Separate abstracts have been prepared. Accelerator activites and computer system/network services are also reported.

  1. Surveys of research in the Chemistry Division, Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Grazis, B.M. (ed.)

    1992-01-01

    Research reports are presented on reactive intermediates in condensed phase (radiation chemistry, photochemistry), electron transfer and energy conversion, photosynthesis and solar energy conversion, metal cluster chemistry, chemical dynamics in gas phase, photoionization-photoelectrons, characterization and reactivity of coal and coal macerals, premium coal sample program, chemical separations, heavy elements coordination chemistry, heavy elements photophysics/photochemistry, f-electron interactions, radiation chemistry of high-level wastes (gas generation in waste tanks), ultrafast molecular electronic devices, and nuclear medicine. Separate abstracts have been prepared. Accelerator activites and computer system/network services are also reported.

  2. General description of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor is a boiling light water cooled, heavy water moderated and vertical pressure tube type reactor with its design optimised for utilisation of thorium for power generation. The core consists of (Th-U233 )O2 and (Th-Pu)O2 fuel with a discharge burn up of 20,000 MWd/Te. This reactor incorporates several features to simplify the design, which eliminate certain systems and components. AHWR design is also optimised for easy replaceability of coolant channels, facilitation of in-service inspection and maintenance and ease of erection. The AHWR design also incorporates several passive systems for performing safety-related functions in the event of an accident. In case of LOCA, emergency coolant is injected through 4 accumulators of 260 m3 capacity directly into the core. Gravity driven water pool of capacity 6000 m3 serves to cool the core for 3 days without operator's intervention. Core submergence, passive containment isolation and passive containment cooling are the added features in AHWR. The paper describes the various process systems, core and fuel design, primary components and safety concepts of AHWR. Plant layout and technical data are also presented. The conceptual design of the reactor has been completed, and the detailed design and development is scheduled for completion in the year 2002. (author)

  3. Goals and requirements for advanced reactor concepts

    International Nuclear Information System (INIS)

    Economic problems and public concerns about safety have lead to a reassessment of current nuclear power plant designs and the development of improved designs or new reactor concepts to better meet the needs of United States utilities. This paper presents a set of goals and requirements, developed by the Idaho National Engineering Laboratory (INEL), to provide a means for evaluating the relative merits of alternate advanced reactor concepts. This set of requirements and goals is intended to be independent of any particular reactor concept, and is predicated on the assumption that nuclear power cannot become viable option until the public is favorable to the use of nuclear power for electric power generation in the United States. Under this assumption, the top level requirements defined for new reactor concepts are (1) public acceptability, (2) acceptable investment risk, (3) competitive life cycle costs, and (4) early deployment. Each of these requirements is supported by several related lower level requirements and design goals that are necessary or desirable to meet the top level requirements

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

    International Nuclear Information System (INIS)

    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

  5. 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

  6. Research Reactor: A Powerhouse of Nuclear Technology in Korea

    International Nuclear Information System (INIS)

    The nuclear era in Korea was opened with the 100 kW KRR-1 of which construction started in 1959. The second research reactor, 2 MW KRR-2, was finished in 1972, around when the first nuclear power plant project was launched. Then the next research reactor HANARO, a 30 MW multi-purpose reactor, started the operation in 1995 and has made the technologies for the research reactor development and utilization matured. The competitiveness of Korean research reactor technology was acknowledged by being selected as the supplier of 5 MW JRTR for Jordan. In addition, Korea is sharing its research reactor technologies with many other countries in the areas of training, engineering service, neutron beam instrument manufacturing, and supply of RI goods production equipment and of the advanced research reactor fuel material. KAERI, as a nuclear research institute, has a well-established R and D infrastructure together with the research reactor operation and utilization technology. It can contribute for the new comers to establish a research reactor facility as well as a research environment using the facility as a tool to build-up their nuclear technology and service capability for their public. (author)

  7. Relocation of a Research Reactor

    International Nuclear Information System (INIS)

    The research reactor RA-8 is placed in Pilcaniyeu Technological Centre (PTC) in the province of Rio Negro, approximately 70 km east from San Carlos de Bariloche city. The first time the reactor went critical was in June 1997 and it is out of operation since March 1999. Due to the intention to relocate the reactor in Bariloche Atomic Centre (BAC) a study has been done in order to assess the technical and economical feasibility. The scope of this study covers the disassembly and transport from PTC to BAC. Relocation of the reactor will reduce costs, time and difficulties in the transport of personnel (operators, researcher, and students) to Pilcaniyeu allowing the performance of nuclear research as well as academic application. The RA-8 is basically a critical facility of enriched uranium with light water as moderator. It is a pool type reactor with low thermal power, maximum 100 W and nominal 10 W. The principal assembly and the associated systems are placed in the reactor hall: consisting of a core, tanks, block, nuclear and conventional instrumentation, moderator system and the neutron source system. Also there is a control room with computers for monitoring together with the safety and control systems. The core is inside two stainless steel concentric tanks communicated with each other that contain water during the operation. The technical feasibility consists in: radiological characterization of the facility, visual inspection of the systems, structures and components, dismantling engineering, mass estimation for disassembly, packing, transport and storage. Economical feasibility has been done in order to evaluate time and costs necessaries for the disassembly and transport from Pilcaniyeu to Bariloche. Regulatory aspects that must be fulfilled were considered in this study. Nothing detectable was found in water samples from pipes of the pumps' well. The systems, structures and components of the RA-8 present in general a good condition of preservation that would

  8. Research Reactors Types and Utilization

    International Nuclear Information System (INIS)

    A nuclear reactor, in gross terms, is a device in which nuclear chain reactions are initiated, controlled, and sustained at a steady rate. The nuclei of fuel heavy atoms (mostly 235U or 239Pu), when struck by a slow neutron, may split into two or more smaller nuclei as fission products,releasing energy and neutrons in a process called nuclear fission. These newly-born fast neutrons then undergo several successive collisions with relatively low atomic mass material, the moderator, to become thermalized or slow. Normal water, heavy water, graphite and beryllium are typical moderators. These neutrons then trigger further fissions, and so on. When this nuclear chain reaction is controlled, the energy released can be used to heat water, produce steam and drive a turbine that generates electricity. The fission process, and hence the energy release, are controlled by the insertion (or extraction) of control rods through the reactor. These rods are strongly neutron absorbents, and thus only enough neutrons to sustain the chain reaction are left in the core. The energy released, mostly in the form of heat, should be continuously removed, to protect the core from damage. The most significant use of nuclear reactors is as an energy source for the generation of electrical power and for power in some military ships. This is usually accomplished by methods that involve using heat from the nuclear reaction to power steam turbines. Research reactors are used for radioisotope production and for beam experiments with free neutrons. Historically, the first use of nuclear reactors was the production of weapons grade plutonium for nuclear weapons. Currently all commercial nuclear reactors are based on nuclear fission. Fusion power is an experimental technology based on nuclear fusion instead of fission.

  9. Design study on the Advanced Recycling Reactor

    International Nuclear Information System (INIS)

    Full text: The design study on the Advanced Recycling Reactor (ARR) has been conducted. This paper presents the pre-conceptual design of the ARR that is a loop-typed sodium cooled reactor with MOX fuel. International Nuclear Recycling Alliance (INRA) takes advantage of international experience and uses the design based on Japan Sodium-cooled Fast Reactor (JSFR) as reference for FOA studies of US DOE, because Japan has conducted R and Ds for the JSFR incorporating thirteen technology enhancements expected to improve safety, enhance economics, and increase reactor reliability. The targets of the ARR are to generate electricity while consuming fuel containing transuranics and to attain cost competitiveness with the similar sized LWRs. INRA proposes 3 evolutions of the ARR; ARR1, a 500 MWe demonstration plant, online in 2025; ARR2, a 1,000 MWe commercial plant, online in 2035; ARR3, a 1,500 MWe full-scale commercial plant, online in 2050. INRA believes the scale-up factor of two is acceptable increase from manufacturing and licensing points of view. Major features of the ARR1 are the following: The reactor core is 70cm high and the volume fraction of fuel is approximately 32%. The conversion ratio of fissile is set up less than 0.6 and the amount of burned TRU is 45-51 kg/TWeh.Decay heat can be removed by natural circulation to improve safety. The primary cooling system consists of two-loop arrangement and the integrated IHX/Pump to improve economics. The steam generator with the straight double-walled tube is used to improve reliability. The ARR1 is co-located with a recycling facility. The overall plant facility arrangement is planned assuming to be constructed and installed in an inland area. The plant consists of a reactor building (including reactor auxiliary facilities and electrical/control systems), a turbine building, and a recycling building. The volume of the reactor building will be approximately 180,000 m3. The capital cost for the ARR1 and the ARR2 are

  10. Advanced reactors: the case for metric design

    International Nuclear Information System (INIS)

    The author argues that DOE should insist that all design specifications for advanced reactors be in the International System of Units (SI) in accordance with the Metric Conversion Act of 1975. Despite a lack of leadership from the federal government, industry has had to move toward conversion in order to compete on world markets. The US is the only major country without a scheduled conversion program. SI avoids the disadvantages of ambiguous names, non-coherent units, multiple units for the same quantity, multiple definitions, as well as barriers to international exchange and marketing and problems in comparing safety and code parameters. With a first step by DOE, the Nuclear Regulatory Commission should add the same requirements to reactor licensing guidelines. 4 references

  11. The Australian Replacement Research Reactor

    Science.gov (United States)

    Kennedy, Shane; Robinson, Robert

    2004-03-01

    The 20-MW Australian Replacement Research Reactor represents possibly the greatest single research infrastructure investment in Australia's history. Construction of the facility has commenced, following award of the construction contract in July 2000, and the construction licence in April 2002. The project includes a large state-of-the-art liquid deuterium cold-neutron source and supermirror guides feeding a large modern guide hall, in which most of the instruments are placed. Alongside the guide hall, there is good provision of laboratory, office and space for support activities. While the facility has "space" for up to 18 instruments, the project has funding for an initial set of 8 instruments, which will be ready when the reactor is fully operational in July 2006. Instrument performance will be competitive with the best research-reactor facilities anywhere, and our goal is to be in the top 3 such facilities worldwide. Staff to lead the design effort and man these instruments have been hired on the international market from leading overseas facilities, and from within Australia, and 7 out of 8 instruments have been specified and costed. At present the instrumentation project carries 10contingency. An extensive dialogue has taken place with the domestic user community and our international peers, via various means including a series of workshops over the last 2 years covering all 8 instruments, emerging areas of application like biology and the earth sciences, and computing infrastructure for the instruments.

  12. Reactor safety research in Sweden

    International Nuclear Information System (INIS)

    Objectives, means and results of Swedish light water reactor safety research during the 1970s are reviewed. The expenditure is about 40 Million Swkr per year excluding industry. Large efforts have been devoted to experimental studies of loss of coolant accidents. Large scale containment response tests for simulated pipe breaks were carried out at the Marviken facility. At Studsvik a method for testing fuel during fast power changes has been developed. Stress corrosion, crack growth and the effect of irradiation on the strength ductility of Zircaloy tube was studied. A method for determining the fracture toughness of pressure vessel steel was developed and it was shown that the fracture toughness was lower than earlier assumed. The release of fission products to reactor water was studied and so was the release, transport and removal of airborne radioactive matter for Swedish BWRs and PWRs. Test methods for iodine filter systems were developed. A system for continuous monitoring of radioactive noble gas stack release was developed for the Ringhals plant. Attention was drawn to the importance of the human factor for reactor safety. Probabilistic methods for risk analysis were applied to the Barsebaeck 2 and Forsmark 3 boiling water reactors. Procedures and working conditions for operator personnel were investigated. (GBn)

  13. Research reactors in Austria - Present situation

    International Nuclear Information System (INIS)

    In the past decades Austria operated three research reactors, the 10 MW ASTRA reactor at Seibersdorf, the 250 kW TRIGA reactor at the Atominstitut and the 1 kW Argonaut reactor at the Technical University in Graz. Since the shut down of the ASTRA on July 31th, 1999 and its immediate decommissioning reactor and the shut down of the Argonaut reactor in Graz on August 31st, 2004 only one reactor remains operational for keeping nuclear competence in Austria which is the 250 kW TRIGA Mark II reactor. (author)

  14. Design Study on the Advanced Recycling Reactor

    International Nuclear Information System (INIS)

    The design study on the Advanced Recycling Reactor (ARR) has been conducted. This paper presents the pre-conceptual design of the ARR that is a loop-typed sodium cooled reactor with MOX fuel. International Nuclear Recycling Alliance (INRA) takes advantage of international experience and uses the design based on Japan Sodium-cooled Fast Reactor (JSFR) as reference for FOA studies of DOE in the U.S., because Japan has conducted R and Ds for the JSFR incorporating thirteen technology enhancements expected to improve safety, enhance economics, and increase reactor reliability. ARR's goal is to generate electricity while consuming fuel containing transuranics and to be cost-competitive with LWRs of similar size. INRA proposes 3 evolutions of the ARR; ARR1, a 500 MWe demonstration plant, online in 2025; ARR2, a 1,000 MWe commercial plant, online in 2035; ARR3, a 1,500 MWe full-scale commercial plant, online in 2050. INRA believes the scale-up factor of two is acceptable increase from manufacturing and licensing points of view. Major features of the ARR1 are the following: The reactor core of 70 cm high is working for a burner of TRU. The conversion ratio of fissile is set up less than 0.6 and the amount of burned TRU is 45-51 kg/TWeh. Decay heat can be removed by natural circulation to improve safety. The primary cooling system consists of two-loop arrangement and the integrated IHX/Pump to improve economics. The steam generator with the straight doublewalled tube is used to improve reliability. The capital cost, the construction schedule and regulatory and licensing schedule are estimated. Furthermore, the technology readiness level and the technology development roadmap are studied and identified to be ready for commercial deployment. (author)

  15. 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)

    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)

  16. Removal heat extraction systems in advanced reactors

    International Nuclear Information System (INIS)

    The two main problems generally attributed to the electricity generation by nuclear power are the security of the facility and the radioactivity of the nuclear wastes, in a way that the only tasks of the European Commission on this matter are to make sure a high level of security in the facilities, as well as an adequate fuel and waste management. In this paper we discuss about the main lines in which the CIEMAT and the Polytechnic University of Valencia are working relative to the study of the passive working systems of the advanced designs reactors. (Author) 24 refs

  17. Outline of advanced boiling water reactor

    International Nuclear Information System (INIS)

    The ABWR (Advanced Boiling Water Reactor) is based on construction and operational experience in Japan, USA and Europe. It was developed jointly by the BWR supplieres, General Electric, Hitachi, and Toshiba, as the next generation BWR for Japan. The Tokyo Electric Power Co. provided leadership and guidance in developing the ABWR, and in combination with five other Japanese electric power companies. The major objectives in developing the ABWR are: 1. Enhanced plant operability, maneuverability and daily load-following capability; 2. Increased plant safety and operating margins; 3. Improved plant availability and capacity factor; 4. Reduced occupational radiation exposure; 5. Reduced radwaste volume, and 6. Reduced plant capital and operating costs. (Liu)

  18. Current status of the world's research reactors

    International Nuclear Information System (INIS)

    Data from the IAEA's Research Reactor Database (RRDB) provides information with respect to the status of the world's research reactors. Some summary data are given. Recent initiatives by the IAEA regarding communications and information flow with respect to research reactors are discussed. Future plans and perspectives are also introduced. (author)

  19. Research nuclear reactor operation management

    International Nuclear Information System (INIS)

    Some aspects of reactor operation management are highlighted. The main mission of the operational staff at a testing reactor is to operate it safely and efficiently, to ensure proper conditions for different research programs implying the use of the reactor. For reaching this aim, there were settled down operating plans for every objective, and procedure and working instructions for staff training were established, both for the start-up and for the safe operation of the reactor. Damages during operation or special situations which can arise, at stop, start-up, maintenance procedures were thoroughly considered. While the technical skill is considered to be the most important quality of the staff, the organising capacity is a must in the operation of any nuclear facility. Staff training aims at gaining both theoretical and practical experience based on standards about staff quality at each work level. 'Plow' sheet has to be carefully done, setting clear the decision responsibility for each person so that everyone's own technical level to be coupled to the problems which implies his responsibility. Possible events which may arise in operation, e.g., criticality, irradiation, contamination, and which do not arise in other fields, have to be carefully studied. One stresses that the management based on technical and scientific arguments have to cover through technical, economical and nuclear safety requirements a series of interlinked subprograms. Every such subprograms is subject to some peculiar demands by the help of which the entire activity field is coordinated. Hence for any subprogram there are established the objectives to be achieved, the applicable regulations, well-defined responsibilities, training of the personnel involved, the material and documentation basis required and activity planning. The following up of positive or negative responses generated by experiments and the information synthesis close the management scope. Important management aspects

  20. NIH Research: Advances in Parkinson's Disease Research

    Science.gov (United States)

    ... of this page please turn JavaScript on. NIH Research: Advances in Parkinson's Disease Research Past Issues / Winter 2014 Table of Contents Story ... Photo courtesy of NIH Advances in Parkinson's Disease Research Story Landis, Ph.D., has been Director of ...

  1. Myrrha, new polyvalent research reactor

    International Nuclear Information System (INIS)

    Myrrha (Multi-purpose hybrid research reactor for high-tech applications) is the first prototype of sub-critical nuclear reactor driven by a particle accelerator (an ADS, accelerator-driven system) at semi-industrial scale (50-100 MW), a safe and easy-to-control technology. In an interview, the manager of this project recalls his curriculum, presents and comments the characteristics of Myrrha, outlines why these ADS are so interesting to produce radio-isotopes, comments the variety of countries and companies involved in this project, outlines the peculiarities of Myrrha in terms of safety and the main technological challenges (a mixing of lead and bismuth for the coolant, control of corrosion by oxygen, an improved reliability based on redundant design and fault tolerance, MOX as fuel). He also evokes competing technologies

  2. Decommissioning of Salaspils Research Reactor

    International Nuclear Information System (INIS)

    The Salaspils Research Reactor (SRR) is out of operation since July 1998 and the decommissioning of SRR was started in 1999 according to the decision of the Government of Latvia. The main decommissioning activities up to 2006 were connected with collecting and conditioning of historical radioactive wastes from different storages outside and inside of reactor hall. The total amount of dismantled materials was about 700 tons, more than 77 tons were conditioned in concrete containers for disposal in repository. The radioactive wastes management technology is discussed in the paper. It was found, that additional efforts must be spent for immobilization of radionuclides in cemented matrix to be comply with the wastes acceptance criteria. The investigations of mechanical stability of water-cement matrix are described and discussed in the paper

  3. Progress with the Australian replacement research reactor

    International Nuclear Information System (INIS)

    Construction of the new Australian Research Reactor, the replacement for the now 46 year old HIFAR research reactor, is approximately 80% completed. Construction of the reactor facility began in April 2002 at ANSTO's Lucas Heights site near Sydney and commissioning is still on track for late 2005. Some details of the progress of construction and licensing and an outline of ANSTO research related to the use of Zircaloy-4 in the core region and reflector vessel of the reactor are given. (author)

  4. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

  5. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    International Nuclear Information System (INIS)

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately

  6. ASME Material Challenges for Advanced Reactor Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Ali Siahpush

    2013-07-01

    This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at higher temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.

  7. Thermochemistry of nuclear fuels in advanced reactors

    International Nuclear Information System (INIS)

    The presence of a large number of elements, accompanied with steep temperature gradient results in dynamic chemistry during nuclear fuel burn-up. Understanding this chemistry is very important for efficient and safe usage of nuclear fuels. The radioactive nature of these fuels puts lot of constraint on regulatory bodies to ensure their accident free operation in the reactors. One of the common aims of advanced fuels is to achieve high burn-up. As burn-up of the fuel increases, chemistry of fission-products becomes increasingly more important. To understand different phenomenon taking place in-pile, many out of-pile experiments are carried out. Extensive studies of thermodynamic properties, phase analysis, thermophysical property evaluation, fuel-fission product clad compatibility are carried out with relevant compounds and simulated fuels (SIMFUEL). All these data are compiled and jointly evaluated using different computational methods to predict fuel behaviour during burn-up. Only when this combined experimental and theoretical information confirms safe operation of the pin, a test pin is prepared and burnt in a test reactor. Every fuel has a different chemistry and different constraints associated with it. In this talk, various thermo-chemical aspects of some of the advanced fuels, mixed carbide, mixed nitride, 'Pu' rich MOX, 'Th' based AHWR fuels and metallic fuels will be discussed. (author)

  8. Novelties in design and construction of the advanced reactors

    International Nuclear Information System (INIS)

    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

  9. Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

    International Nuclear Information System (INIS)

    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.

  10. 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.

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

    International Nuclear Information System (INIS)

    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.

  12. Analysis of higher power research reactors' parameters

    International Nuclear Information System (INIS)

    The objective of this monograph was to analyze and compare parameters of different types of research reactors having higher power. This analysis could be used for decision making and choice of a reactor which could possibly replace the existing ageing RA reactor in Vinca. Present experimental and irradiation needs are taken into account together with the existing reactors operated in our country, RB and TRIGA reactor

  13. Operating experiences of the research reactors

    International Nuclear Information System (INIS)

    Nuclear research reactors are devices of wide importance, being used for different scientific research tasks, for testing and improving reactor systems and components, for the production of radioisotopes, for the purposes of defence, for staff training and for other purposes. There are three research reactors in Yugoslavia: RA, RB and TRIGA. Reactors RA and RB at the 'Boris Kidric' Institute of Nuclear Sciences are of heavy water type power being 6500 and 10 kW, and maximum thermal neutron flux of 1014 and 1011(n/cm2s), respectively. TRIGA reactor at the 'Jozef Stefan' Institute in Ljubljana is of 250 kW power and maximum thermal neutron flux of 1013(n/cm2s). Reactors RA and RB use soviet fuel in the form of uranium dioxide (80% enriched) and metallic uranium (2%). Besides, RB reactor operates with natural uranium too. TRIGA reactor uses american uranium fuel 70% and 20% enriched, uranium being mixed homogeneously with moderator (ZrH). Experiences in handling and controlling the fuel before irradiation in the reactor, in reactor and after it are numerous and valuable, involving either the commercial arrangements with foreign producers, or optimal burn up in reactor or fuel treatment after the reactor irradiation. Twenty years of operating experience of these reactors have great importance especially having in mind the number of trained staff. Maintenance of reactors systems and fluids in continuous operation is valuable experience from the point of view of water reactor utilization. The case of the RA reactor primary cycle cobalt decontamination and other events connected with nuclear and radiation security for all three reactors are also specially emphasized. Owing to our research reactors, numerous theoretical, numerical and experimental methods are developed for nuclear and other analyses and design of research and power reactors,as well as methods for control and protection of radiation. (author)

  14. Training and Certification of Research Reactor Personnel

    International Nuclear Information System (INIS)

    The safe operation of a research reactor requires that reactor personnel be fully trained and certified by the relevant authorities. Reactor operators at PUSPATI TRIGA Reactor underwent extensive training and are certified, ever since the reactor first started its operation in 1982. With the emphasis on enhancing reactor safety in recent years, reactor operator training and certification have also evolved. This paper discusses the changes that have to be implemented and the challenges encountered in developing a new training programme to be in line with the national standards. (author)

  15. Nuclear research reactors activities in INVAP

    International Nuclear Information System (INIS)

    This presentation describes the different activities in the research reactor field that are being carried out by INVAP. INVAP is presently involved in the design of three new research reactors in three different countries. The RA-10 is a multipurpose reactor, in Argentina, planned as a replacement for the RA-3 reactor. INVAP was contracted by CNEA for carrying out the preliminary engineering for this reactor, and has recently been contracted by CNEA for the detailed engineering. CNEA groups are strongly involved in the design of this reactor. The RMB is a multipurpose reactor, planned by CNEN from Brazil. CNEN, through REDETEC, has contracted INVAP to carry out the preliminary engineering for this reactor. As the user requirements for RA-10 and RMB are very similar, an agreement was signed between Argentina and Brasil governments to cooperate in these two projects. The agreement included that both reactors would use the OPAL reactor in Australia, design and built by INVAP, as a reference reactor. INVAP has also designed the LPRR reactor for KACST in Saudi Arabia. The LPRR is a 30 kw reactor for educational purposes. KACST initially contracted INVAP for the engineering for this reactor and has recently signed the contract with INVAP for building the reactor. General details of these three reactors will be presented

  16. Nuclear research reactors activities in INVAP

    Energy Technology Data Exchange (ETDEWEB)

    Ordonez, Juan Pablo [INVAP, Bariloche (Argentina)

    2013-07-01

    This presentation describes the different activities in the research reactor field that are being carried out by INVAP. INVAP is presently involved in the design of three new research reactors in three different countries. The RA-10 is a multipurpose reactor, in Argentina, planned as a replacement for the RA-3 reactor. INVAP was contracted by CNEA for carrying out the preliminary engineering for this reactor, and has recently been contracted by CNEA for the detailed engineering. CNEA groups are strongly involved in the design of this reactor. The RMB is a multipurpose reactor, planned by CNEN from Brazil. CNEN, through REDETEC, has contracted INVAP to carry out the preliminary engineering for this reactor. As the user requirements for RA-10 and RMB are very similar, an agreement was signed between Argentina and Brasil governments to cooperate in these two projects. The agreement included that both reactors would use the OPAL reactor in Australia, design and built by INVAP, as a reference reactor. INVAP has also designed the LPRR reactor for KACST in Saudi Arabia. The LPRR is a 30 kw reactor for educational purposes. KACST initially contracted INVAP for the engineering for this reactor and has recently signed the contract with INVAP for building the reactor. General details of these three reactors will be presented.

  17. Rationalization and future planning for AECL's research reactor capability

    International Nuclear Information System (INIS)

    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

  18. Status of advanced light water reactor designs 2004

    International Nuclear Information System (INIS)

    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

  19. Reliability studies in research reactors

    International Nuclear Information System (INIS)

    Fault trees and event trees are widely used in industry to model and to evaluate the reliability of safety systems. Detailed analyzes in nuclear installations require the combination of these two techniques. This study uses the methods of FT (Fault Tree) and ET (Event Tree) to accomplish the PSA (Probabilistic Safety Assessment) in research reactors. According to IAEA (lnternational Atomic Energy Agency), the PSA is divided into Level 1, Level 2 and Level 3. At the Level 1, conceptually, the security systems perform to prevent the occurrence of accidents, At the Level 2, once accidents happened, this Level seeks to minimize consequences, known as stage management of accident, and at Level 3 accident impacts are determined. This study focuses on analyzing the Level 1, and searching through the acquisition of knowledge, the consolidation of methodologies for future reliability studies. The Greek Research Reactor, GRR-1, is a case example. The LOCA (Loss of Coolant Accident) was chosen as the initiating event and from it, using ET, possible accidental sequences were developed, which could lead damage to the core. Moreover, for each of affected systems, probabilities of each event top of FT were developed and evaluated in possible accidental sequences. Also, the estimates of importance measures for basic events are presented in this work. The studies of this research were conducted using a commercial computational tool SAPHIRE. Additionally, achieved results thus were considered satisfactory for the performance or the failure of analyzed systems. (author)

  20. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  1. Education and Training Programme at the Research Reactor TRIGA Mainz

    Energy Technology Data Exchange (ETDEWEB)

    Hampel, Gabriele; Eberhardt, Klaus [University of Mainz, Institute for Nuclear Chemistry, D-55099 Mainz (Germany)

    2011-07-01

    Education and training are important elements for the future of nuclear science, technology and safety. Fields of interest include high- technology applications in nuclear techniques and neutron sources, advances in the areas of power reactor safety, establishing the scientific basis of new reactors, training of personnel needed to operate, maintain, regulate and improve reactors or other facilities associated with nuclear power. Also, creating a knowledgeable public through education usually means less opposition and more support. Education and training for safeguards, operators, researchers and quality programmes (calibration services, etc.) are one of the main utilisations of TRIGA research reactors. Use of a reactor as a training tool for university students studying nuclear engineering and/or physics, where there is a growing demand at European Universities, is of vital importance. In particular, the TRIGA Mark II reactor, located at the University of Mainz, one of the largest universities in Germany, offers a broad range of nuclear-related courses for training and education. (author)

  2. Education and Training Programme at the Research Reactor TRIGA Mainz

    International Nuclear Information System (INIS)

    Education and training are important elements for the future of nuclear science, technology and safety. Fields of interest include high- technology applications in nuclear techniques and neutron sources, advances in the areas of power reactor safety, establishing the scientific basis of new reactors, training of personnel needed to operate, maintain, regulate and improve reactors or other facilities associated with nuclear power. Also, creating a knowledgeable public through education usually means less opposition and more support. Education and training for safeguards, operators, researchers and quality programmes (calibration services, etc.) are one of the main utilisations of TRIGA research reactors. Use of a reactor as a training tool for university students studying nuclear engineering and/or physics, where there is a growing demand at European Universities, is of vital importance. In particular, the TRIGA Mark II reactor, located at the University of Mainz, one of the largest universities in Germany, offers a broad range of nuclear-related courses for training and education. (author)

  3. Halden Reactor Project Workshop: Understanding Advanced Instrumentation and Controls Issues

    International Nuclear Information System (INIS)

    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

  4. Seismic research on graphite reactor core

    International Nuclear Information System (INIS)

    Background: Reactors with graphite core structure include production reactor, water-cooled graphite reactor, gas-cooled reactor, high-temperature gas-cooled reactor and so on. Multi-body graphite core structure has nonlinear response under seismic excitation, which is different from the response of general civil structure, metal connection structure or bolted structure. Purpose: In order to provide references for the designing and construction of HTR-PM. This paper reviews the history of reactor seismic research evaluation from certain countries, and summarizes the research methods and research results. Methods: By comparing the methods adopted in different gas-cooled reactor cores, inspiration for our own HTR seismic research was achieved. Results and Conclusions: In this paper, the research ideas of graphite core seismic during the process of designing, constructing and operating HTR-10 are expounded. Also the project progress of HTR-PM and the research on side reflection with the theory of similarity is introduced. (authors)

  5. New Research Reactor Project in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Sangik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    New research reactor project in Korea aims to increase self-sufficiency in terms of medical and industrial radioisotope supply, to enlarge the supply of NTD silicon doping and to make up the advanced technology related to research reactor. As a major national project for nuclear science and engineering in Korea, this project has been officially started on Apr. 2012 funded by the Government and Local Governments, Kijang-Gun and Busan City, that accommodate this facility in their land. It has five years project schedule from design to commissioning just before starting the normal operation in 2017. KAERI, who is the owner contracted with the Government, is doing the design by itself based on our own knowledge and experiences from the KRR-1, KRR-2 and HANARO. The reactor is be composed of 8.0 g/cc U-Mo fuel, which is the First-of-a-kind application in world, and will be enable of operating over 300 days per year and 60% high burn-up pertinently to produce the required neutron flux. It is expected that the construction permit application will be submitted to regulatory body by the first half of 2014 and the first criticality by 2017.

  6. Systemization of Design and Analysis Technology for Advanced Reactor

    International Nuclear Information System (INIS)

    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

  7. Fuel rod bundles proposed for advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    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 (ACRTM-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)

  8. The Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Todd R. Allen; Collin J. Knight; Jeff B. Benson; Frances M. Marshall; Mitchell K. Meyer; Mary Catherine Thelen

    2011-08-01

    In 2007, the Advanced Test Reactor (ATR), located at Idaho National Laboratory (INL), was designated by the Department of Energy (DOE) 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 approved researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide those 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, obtained access to additional PIE equipment, taken steps to enable the most advanced post-irradiation analysis possible, and initiated an educational program and digital learning library to help potential users better understand the critical issues in reactor technology and how a test reactor facility could be used to address this critical research. 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 invited universities to nominate their capability to become part of a broader user facility. Any university is eligible to self-nominate. Any nomination is then peer reviewed to ensure that the addition of the university facilities adds useful capability to the NSUF. Once added to the NSUF team, the university capability is then integral to the NSUF operations and is available to all users via the proposal process. So far, six universities 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 university capabilities, 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. The current NSUF partners are

  9. ALWR - VTT's technology programme on advanced light water reactors

    International Nuclear Information System (INIS)

    In January 1998, VTT Energy launched a four year research programme 'Advanced Light Water Reactors' (ALWR). The research programme strives to intensify co operation between different Finnish organisations as well as to make better use of international R and D efforts on ALWR concepts. Some projects of the programme focus on use of new computational tools for the design and safety analysis of ALWRs. Improved computer simulation methods and models are necessary for the analysis of passive safety systems. Some projects address development and assessment of new technical solutions in ALWR. The projects are carried out in close co operation with the Finnish power companies and are often part of reactor vendors international development programmes. Another key element of the programme is training of new nuclear experts and providing possibilities for continuing education of the current staff. (author)

  10. Design Requirements of an Advanced HANARO Reactor Core Cooling System

    International Nuclear Information System (INIS)

    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

  11. Study of Second Phase Particles and Fe content in Zr Alloys Using the Advanced Photon Source at Argonne

    International Nuclear Information System (INIS)

    We have conducted a study of second phase particles and matrix alloying element concentrations in zirconium alloys using synchrotron radiation from the Advanced Photon Source (APS) at Argonne National Laboratory. The high flux of synchrotron radiation delivered at the 2BM beamline compared to conventional x-ray generators, enables the detection of very small precipitate volume fractions. We detected the standard C14 hcp Zr(Cr,Fe)2 precipitates, (the stable second phase in Zircaloy-4) in the bulk material at a cumulative annealing parameter as low as 10-20 h, and we followed the kinetics of precipitation and growth as a function of the cumulative annealing parameter (CAP) in the range 10-22 (quench) to 10-16 h. In addition, the unique combination of spatial resolution and elemental sensitivity of the 2ID-D/E microbeam line at the Advanced Photon Source at Argonne (APS) allows study of the alloying element concentrations at ppm levels in an area as small as 0.2 mm. We used x-ray fluorescence induced by this sub-micron x-ray beam to determine the concentration of these alloying elements in the matrix as a function of alloy type and thermal history. We discuss these results and the potential of synchrotron radiation-based techniques for studying zirconium alloys

  12. Operation experiences and utilizations of Korean research reactors

    International Nuclear Information System (INIS)

    There are two TRIGA research reactors, MARK-II of 250 KW and MARK-III of 2,000 KW, currently in operation at the Korea Advanced Energy Research Institute in Seoul, Korea. The major utilizations of research reactors in Korea include : (1) radioisotope production and activation analysis, (2) neutron research, and (3) training. The accumulated operation time as of the end of 1985 has totaled 33,084 hours for MARK-II, and 28,807 hours for MARK-III. During this period, 24,983 cases of sample irradiations have been carried out, while the number of trainees reached 2,788. Recently, the KAERI undertook a plan to build a 30 MW multipurpose research reactors at Daeduk research center. At present, this new reactor is in the preliminary design stage. (author)

  13. Safe Operation of Research Reactors in Germany

    International Nuclear Information System (INIS)

    In Germany, experience was gained in the field of safe operation of research reactors during the last five decades. In this time, in total 46 research reactors were built and operated safely. Concerning the design, there is, or has been, a very broad range of different types of research reactors. The variety of facilities includes large pool or tank reactors with a thermal power of several tens of megawatt as well as small educational reactors with a negligible thermal power and critical assemblies. At present, 8 research reactors are still in operation. The other facilities are permanently shutdown, in decommissioning or have already been dismantled completely and released from regulatory control. In this paper, four selected facilities still being operated are presented as examples for safe operation of research reactors in Germany, including especially a description of the safety reviews and safety upgrades for the older facilities. (author)

  14. Advances in neutron scattering research

    International Nuclear Information System (INIS)

    This issue of the Supplement to Journal of the Physical Society of Japan collects invited and contributed papers from the first International Symposium on Advanced Science Research (ASR-2000) 'Advances in Neutron Scattering Research'. The 182 of the presented papers are indexed individually. (J.P.N.)

  15. Application of research reactors for radiation education

    International Nuclear Information System (INIS)

    Nuclear research Reactors are, as well as being necessary for research purposes, indispensable educational tools for a country whose electric power resources are strongly dependent on nuclear energy. Both large and small research reactors are available, but small ones are highly useful from the viewpoint of radiation education. This paper oders a brief review of how small research reactors can, and must, be used for radiation education for high school students, college and graduate students, as well as for the public. (author)

  16. Advanced nuclear reactor public opinion project

    Energy Technology Data Exchange (ETDEWEB)

    Benson, B.

    1991-07-25

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

  17. Advanced nuclear reactor public opinion project

    International Nuclear Information System (INIS)

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions

  18. Technical meeting - First research co-ordination meeting (RCM) of the co-ordinated research project (CRP) on 'Studies of advanced reactor technology options for effective incineration of radioactive waste'. Working material

    International Nuclear Information System (INIS)

    Different strategies for dealing with nuclear waste are being followed by various countries because of their geologic situations and their views on nuclear energy, reprocessing and non-proliferation. Partitioning and Transmutation (P and T) offers potential alternative paths that would, if successful, essentially eliminate plutonium, higher actinides, and environmentally hazardous long-lived fission products from the waste stream destined for permanent storage. One possibility, for which renewed interest has been triggered by proposals made around 1994 by C. Rubbia at CERN, to implement P and T is linked to the development of sub-critical hybrid systems that combine a powerful proton accelerator to produce spallation neutrons with a sub-critical core driven by this source to a relatively high fission power. Accelerator Driven Systems (ADS) can also be envisaged for power production. The efficiency of both critical reactors and ADS for the transmutation and incineration of nuclear waste is strongly related to the utilization of so-called dedicated fuels (solid fuelled reactors). In the ideal case, these fuels should consist of pure TRUs without fertile materials, such as or 232-Th, to achieve highest incineration/transmutation rates. The IAEA Coordinated Research Program (CRP) on 'Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste' has received 17 proposals from various R and D institutions in 13 Member States and the BC (JRC). The overall objective of the CRP is to perform R and D tasks contributing towards the proof of practicality for long-lived waste transmutation. For a sound assessment of the transient and accident behaviour, the neutron kinetics and dynamics have to be qualified, especially as the margins for the safety relevant neutronics parameters are becoming small in a 'dedicated' transmuter. The CRP will integrate benchmarking of transient/accident simulation codes focussing on the phenomena and effects

  19. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented

  20. Fuel behavior in advanced water reactors

    International Nuclear Information System (INIS)

    Fuel rod behavior of advanced pressurized water reactors under steady state conditions has been investigated in this study. System-80+ and Westinghouse Vantage-5 fuels have been considered as advanced pressurized water reactor fuels to be analyzed. The purpose of this study is to analyze the sensitivity of ditferent models and the effect of selected design parameters on the overall fuel behavior. FRAPCON-II computer code has been used for the analyses. Different modelling options of FRAPCON-II have also been considered in these analyses. Analyses have been performed in two main parts. In the first part, effects of operating conditions on fuel behavior have been investigated. First, fuel rod response under normal operating conditions has been analyzed. Then, fuel rod response to different fuel ratings has been calculated. In the second part, in order to estimate the effect of design parameters on fuel behavior, parametric analyses have been performed. In this part, the effects of initial gap thickness, as fabricated fuel density, and initial fill gas pressure on fuel behavior have been analyzed. The computations showed that both of the fuel rods used in this study operate within the safety limits. However, FRAPCON-II modelling options have been resulted in different behavior due to their modelling characteristics. Hence, with the absence of experimental data, it is difficult to make assesment for the best fuel parameters. It is also difficult to estimate error associated with the results. To improve the performance of the code, it is necessary to develop better experimental correlations for material properties in order to analyze the eftect ot considerably different design parameters rather than nominal rod parameters

  1. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    The purpose of this technical paper is to provide status of the United State domestic Research Reactor Infrastructure (RRI) Program at the Idaho National Laboratory. This paper states the purpose of the program, lists the universities operating TRIGA reactors that are supported by the program, identifies anticipated fresh fuel needs for the reactor facilities, discusses spent fuel activities associated with the program, and addresses successes and planned activities for the program. (author)

  2. Human-health effects of radium: an epidemiolgic perspective of research at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Stebbings, J.H.

    1982-01-01

    The topic of health effects of radium has recently been considerably broadened by the identification of multiple myeloma as a specific outcome of bone-seeking radionuclides, and by evidence that the incidence of breast cancer may be significantly increased by radium exposure. All soft-tissue tumors are now suspect, especially leukemias. Concepts of dose-response need to be broadened to include the concept of risk factors, or, if one prefers, of susceptible subgroups. Biological factors relating to radium uptake and retention require study, as do risk factors modifying risk of both the clasical tumors, osteosarcoma and nasal sinus/mastoid, and the more recently suspect soft-tissue tumors. The history, organization, and current research activities in epidemiology at Argonne National Laboratory are described, and findings of the last decade and a half reviewed. Plans for future research are briefly discussed.

  3. Human health effects of radium: an epidemiologic perspective of research at Argonne National Laboratory

    International Nuclear Information System (INIS)

    The topic of health effects of radium has recently been considerably broadened by the identification of multiple myeloma as a specific outcome of bone-seeking radionuclides, and by evidence that the incidence of breast cancer may be significantly increased by radium exposure. All soft-tissue tumors are now suspect, especially leukemias. Concepts of dose-response need to be broadened to include the concept of risk factors, or, if one prefers, of susceptible subgroups. Biological factors relating to radium uptake and retention require study, as do risk factors modifying risk of both the classical tumors, osteosarcoma and nasal sinus/mastoid, and the more recently suspect soft-tissue tumors. The history, organization, and current research activities in epidemiology at Argonne National Laboratory are described, and findings of the last decade and a half reviewed. Plans for future research are briefly discussed

  4. Human-health effects of radium: an epidemiolgic perspective of research at Argonne National Laboratory

    International Nuclear Information System (INIS)

    The topic of health effects of radium has recently been considerably broadened by the identification of multiple myeloma as a specific outcome of bone-seeking radionuclides, and by evidence that the incidence of breast cancer may be significantly increased by radium exposure. All soft-tissue tumors are now suspect, especially leukemias. Concepts of dose-response need to be broadened to include the concept of risk factors, or, if one prefers, of susceptible subgroups. Biological factors relating to radium uptake and retention require study, as do risk factors modifying risk of both the clasical tumors, osteosarcoma and nasal sinus/mastoid, and the more recently suspect soft-tissue tumors. The history, organization, and current research activities in epidemiology at Argonne National Laboratory are described, and findings of the last decade and a half reviewed. Plans for future research are briefly discussed

  5. Upgrading I and C for the Es Salam research reactor

    International Nuclear Information System (INIS)

    The Es Salam is a multi-purpose research reactor intended for the production of the radio elements, material tests, education and training. It serves also as a source for neutron beams used by chemists, biologists, metallurgists and physicists for fundamental research and applications. The main task of the Es Salam Reactor Operation Division is to ensure a safe and reliable operation of the reactor. With this objective in hand, the Operation Division is responsible for updating the safety measures and conditions in the installation for the reactor equipments and systems. Due to the increased demand for experiences and to the ageing effects, modification and modernization of some safety items become necessary in comparison with state of the art installations. Furthermore, the technological advances and the development and introduction of new instruments, components and systems increased the need for modification. In this paper, we present the new I and C system which will be used for the Es Salam research reactor

  6. Euratom research supporting reactor safety

    International Nuclear Information System (INIS)

    This paper focuses on the way the Fifth Framework Programme (FP5) of the European Atomic Energy Community (Euratom) (1998-2002) contributed to the overall effort supporting LWR safety and on the prospects of FP6 (2002-2006) to achieve the additional goal of contributing to the creation of the European Research Area (ERA). The objectives of the FP5 Community research are discussed, topic per topic, with reference to the relevant projects contributing to their achievement while emphasising the role and expectations of the end-users. 71 research projects have been carried out in FP5 with a total cost of 85.4 million Euro, out of which 43.5 million Euro are contributed by the EU. Economic aspects and needs for nuclear knowledge management throughout the EU-25 countries are briefly discussed as well as future Euratom research needs and nuclear stakeholders' interests. Implementation aspects and prospects of Euratom FP6 are discussed bearing in mind the challenging additional objective. Very ambitious S/T actions have been undertaken supporting reactor safety not only by their technical achievements but also by their structuring effect in the European research picture. (authors)

  7. Update on fuel fabrication development and testing at Argonne National Laboratory

    International Nuclear Information System (INIS)

    In its effort to develop research reactor fuel with a high fissile loading, Argonne National Laboratory has continued its advanced fuel development efforts. Monolithic fuel, where the fuel is in the form of a single fuel foil, is being developed as the ultimate in fuel loading capacity. Work has been done on different monolithic fabrication methods that have resulted in process refinements. Effort is also underway to develop a uranium-molybdenum dispersion fuel plate that will be resistant to the irradiation shortcomings noted in previous tests. Alloying additions to the aluminum matrix are being investigated. These fuels are being fabricated for use in irradiation experiments scheduled for insertion in 2005. (author)

  8. RB research reactor Safety Report

    International Nuclear Information System (INIS)

    This RB reactor safety report is a revised and improved version of the Safety report written in 1962. It contains descriptions of: reactor building, reactor hall, control room, laboratories, reactor components, reactor control system, heavy water loop, neutron source, safety system, dosimetry system, alarm system, neutron converter, experimental channels. Safety aspects of the reactor operation include analyses of accident causes, errors during operation, measures for preventing uncontrolled activity changes, analysis of the maximum possible accident in case of different core configurations with natural uranium, slightly and highly enriched fuel; influence of possible seismic events

  9. Korea Research Reactor -1 and 2 Decommissioning Project in Korea

    International Nuclear Information System (INIS)

    Korea Research Reactor 1 (KRR-1), the first research reactor in Korea, has been operated since 1962, and the second one, Korea Research Reactor 2 (KRR-2) since 1972. The operation of both of them was phased out in 1995 due to their lifetime and operation of the new and more powerful research reactor, HANARO (High-flux Advanced Neutron Application Reactor; 30MW). Both are TRIGA Pool type reactors in which the cores are small self-contained units sitting in tanks filled with cooling water. The KRR-1 is a TRIGA Mark II, which could operate at a level of up to 250 kW. The second one, the KRR-2 is a TRIGA Mark III, which could operate at a level of up 2,000 kW. The decontamination and decommissioning (D and D) project of these two research reactors, the first D and D project in Korea, was started in January 1997 and will be completed to stage 3 by 2008. The aim of this decommissioning program is to decommission the KRR-1 and 2 reactors and to decontaminate the residual building structure s and the site to release them as unrestricted areas. KAERI (Korea Atomic Energy Research Institute) submitted the decommissioning plan and the environmental impact assessment reports to the Ministry of Science and Technology (MOST) for the license in December 1998, and was approved in November 2000

  10. Korea Research Reactor -1 & 2 Decommissioning Project in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. K.; Chung, U. S.; Jung, K. J.; Park, J. H.

    2003-02-24

    Korea Research Reactor 1 (KRR-1), the first research reactor in Korea, has been operated since 1962, and the second one, Korea Research Reactor 2 (KRR-2) since 1972. The operation of both of them was phased out in 1995 due to their lifetime and operation of the new and more powerful research reactor, HANARO (High-flux Advanced Neutron Application Reactor; 30MW). Both are TRIGA Pool type reactors in which the cores are small self-contained units sitting in tanks filled with cooling water. The KRR-1 is a TRIGA Mark II, which could operate at a level of up to 250 kW. The second one, the KRR-2 is a TRIGA Mark III, which could operate at a level of up 2,000 kW. The decontamination and decommissioning (D & D) project of these two research reactors, the first D & D project in Korea, was started in January 1997 and will be completed to stage 3 by 2008. The aim of this decommissioning program is to decommission the KRR-1 & 2 reactors and to decontaminate the residual building structure s and the site to release them as unrestricted areas. KAERI (Korea Atomic Energy Research Institute) submitted the decommissioning plan and the environmental impact assessment reports to the Ministry of Science and Technology (MOST) for the license in December 1998, and was approved in November 2000.

  11. Role of advanced reactors in further nuclear power development

    International Nuclear Information System (INIS)

    As a part of the national long-term nuclear R and D program launched in 1992, an endeavor has been made in Korea to develop advanced nuclear reactor systems with significantly enhanced safety and economics from those of the current generation nuclear power plants. The advanced PWR nuclear reactor systems under development in Korea include 1300 MWe Korean Next Generation Reactor (KNGR), 330 MWt Integral Type System Integrated Modular Advanced Reactor (SMART) for nuclear cogeneration, and 330 MWe Korea Advanced Liquid Metal Reactor (KALIMER) in addition to the evolutionary enhancement of the 1000 MWt KSNPP (Korea Standard Nuclear Power Plant). Three point design philosophy has been adopted for the development of the advanced reactors in Korea : enhancements on safety, economics and public acceptance of nuclear power. To enhance the safety of the advanced reactor systems, a strategy has been adopted to employ advanced design features as well as the passive safety design features. Economically viable design concepts also have been implemented in the evolutionary KSNPP, KNGR, and the SMART development. Economic competitiveness against the fossil plants also has been set as a major objective of the ALWR development program in Korea. These safer and more economical advanced reactors will better promote the public acceptance of the commercial use of the nuclear power and thus could be utilized to meet the forecasted national energy need in the early 21st century. International cooperation in the areas of ALWR development as well as improving public acceptance of the nuclear power is required. (author)

  12. Potential for new societal contributions from the advanced test reactor

    International Nuclear Information System (INIS)

    The mission of the Advanced Test Reactor (ATR) at Idaho National Engineering Laboratory is to study the effects of intense radiation on materials and fuels and to produce radioisotopes for the U.S. Department of Energy (DOE) for government and commercial applications. The purpose of this paper is to explore the potential benefits to society from these available neutrons. The ATR is a 250-MW(thermal) light water reactor highly enriched uranium in plate-type fuel. The ATR uses a combination of hafnium control drums and shim rods to adjust power and hold flux distortion to a minimum. The different quadrants of the ATR can be operated at different power levels to meet a variety of mission requirements. Irradiation positions are available at various locations throughout the core and beryllium reflector. A summary of the flux levels at various ATR reflector and loop positions. is given. These fluxes are maintained with a relatively constant axial flux profile throughout cycles that last 35 to 42 days. These neutrons can be used for testing and irradiation programs that support commercial reactor license extension, advanced fuel development, materials effects studies, failure cause/effect studies, coolant chemistry evaluations, prototype testing programs, isotope production, and basic research. Radioisotope production falls into three categories: medical, industrial, and research. In summary, the ATR is a unique, high-power test reactor capable of supporting the current DOE mission and producing radioisotopes. Space available for radioisotope production and fuels or materials testing will increase by 44% in 1994, improving DOE's ability to support national needs in health care, industry, and research

  13. Strategic planning for research reactors. Guidance for reactor managers

    International Nuclear Information System (INIS)

    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

  14. Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Foyto, L [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Kutikkad, K [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; McKibben, J C [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Peters, N. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Stevens, J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2012-09-01

    This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.

  15. Current activities at the MIT research reactor

    International Nuclear Information System (INIS)

    The MIT Research Reactor (MITR) is a 5 MW nuclear research reactor that is owned and operated by the Massachusetts Institute of Technology to further its educational and research goals at both the undergraduate and graduate level. The facility (MITR-II) uses finned, aluminum-clad, plate-type fuel that is cooled and moderated by light water and reflected by heavy water. This paper provides an overview of current activities at the MITR including: (1) The current operating license will expire in August 1999. A decision has been made to pursue a power upgrade to the maximum level (6-7 MW) that can be safely supported by the existing heat removal equipment. Preparation of relicensing documents and results of thermal hydraulic studies are reviewed. (2) The status of an on-going phase-I clinical trial of BNCT for both glioblastoma multiform and metastatic melanoma will be reported. (3) A fission converter facility has been designed for advanced BNCT clinical trials and for routine therapy. This facility will provide a high quality epithermal neutron beam which is capable of treating a patient in a few minutes. Construction of the facility is currently in progress. The facility's design is summarized. (4) A recent study that was completed at the MIT-II using NAA is reported. This study entailed evaluation of the air quality in Upstate New York from October 1991 through September 1993. (5) A number of unique experimental water loop facilities for the study of light water power reactor coolant chemistry have been installed and operated in the MITR-II. The capabilities and the research objectives addressed by these facilities are summarized. (author)

  16. Current activities at the MIT research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Lin Wen; Bernard, John A.; Harling, Otto K.; Kohse, Gordon E.; Olmez, Ilhan [MIT, Cambridge (United States)

    1998-07-01

    The MIT Research Reactor (MITR) is a 5 MW nuclear research reactor that is owned and operated by the Massachusetts Institute of Technology to further its educational and research goals at both the undergraduate and graduate level. The facility (MITR-II) uses finned, aluminum-clad, plate-type fuel that is cooled and moderated by light water and reflected by heavy water. This paper provides an overview of current activities at the MITR including: (1) The current operating license will expire in August 1999. A decision has been made to pursue a power upgrade to the maximum level (6-7 MW) that can be safely supported by the existing heat removal equipment. Preparation of relicensing documents and results of thermal hydraulic studies are reviewed. (2) The status of an on-going phase-I clinical trial of BNCT for both glioblastoma multiform and metastatic melanoma will be reported. (3) A fission converter facility has been designed for advanced BNCT clinical trials and for routine therapy. This facility will provide a high quality epithermal neutron beam which is capable of treating a patient in a few minutes. Construction of the facility is currently in progress. The facility's design is summarized. (4) A recent study that was completed at the MIT-II using NAA is reported. This study entailed evaluation of the air quality in Upstate New York from October 1991 through September 1993. (5) A number of unique experimental water loop facilities for the study of light water power reactor coolant chemistry have been installed and operated in the MITR-II. The capabilities and the research objectives addressed by these facilities are summarized. (author)

  17. A Joint Report on PSA for New and Advanced Reactors

    International Nuclear Information System (INIS)

    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

  18. Light water reactor safety research project

    International Nuclear Information System (INIS)

    The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

  19. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    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.

  20. Safety of research reactors - A regulator's perspective

    International Nuclear Information System (INIS)

    Due to historical reasons research reactors have received less regulatory attention in the world than nuclear power plants. This has given rise to several safety issues which, if not addressed immediately, may result in an undesirable situation. However, in Pakistan, research reactors and power reactors have received due attention from the regulatory authority. The Pakistan Research Reactor-1 has been under regulatory surveillance since 1965, the year of its commissioning. The second reactor has also undergone all the safety reviews and checks mandated by the licensing procedures. A brief description of the regulatory framework, the several safety reviews carried out have been briefly described in this paper. Significant activities of the regulatory authority have also been described in verifying the safety of research reactors in Pakistan along with the future activities. The views of the Pakistani regulatory authority on the specific issues identified by the IAEA have been presented along with specific recommendations to the IAEA. We are of the opinion that there are more Member States operating nuclear research reactors than nuclear power plants. Therefore, there should be more emphasis on the research reactor safety, which somehow has not been the case. In several recommendations made to the IAEA on the specific safety issues the emphasis has been, in general, to have a similar documentation and approach for maintaining and verifying operational safety at research reactors as is currently available for nuclear power reactors and may be planned for nuclear fuel cycle facilities. (author)

  1. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

    In 1985, the International Atomic Energy Agency (IAEA) published a report titled 'Status of Liquid Metal Cooled Fast Breeder Reactors' (Technical Reports Series No. 246). The report was a general review of the status of fast reactor development at that time, covering some aspects of design and operation and reviewing experience from the earliest days. It summarized the programmes and plans in all countries which were pursuing the development of fast reactors. In 1999, the IAEA published a follow-up report titled 'Status of Liquid Metal Cooled Fast Reactor Technology' (IAEA-TECDOC-1083), necessitated by the substantial advances in fast reactor technology development and changes in the economic and regulatory environment which took place during the period of 1985-1998. Chief among these were the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at a high burnup and the launch of new fast reactor programmes by some additional Member States. In 2006, the Technical Working Group on Fast Reactors (TWG-FR) identified the need to update its past publications and recommended the preparation of a new status report on fast reactor technology. The present status report intends to provide comprehensive and detailed information on the technology of fast neutron reactors. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction, operation and decommissioning; various areas of research and development; engineering; safety; and national strategies and public acceptance of fast reactors.

  2. Overview of basic and applied research on battery systems at Argonne

    Energy Technology Data Exchange (ETDEWEB)

    Nevitt, M. V.

    1979-01-01

    The need for a basic understanding of the ion transport and related effects that are observed under the unique physical and electrochemical conditions occurring in high-temperature, high-performance batteries is pointed out. Such effects include those that are typical of transport in bulk materials such as liquid and solid electrolytes and the less well understood effects observed in migration in and across the interfacial zones existing around electrodes. The basic and applied studies at Argonne National Laboratory, centered in part around the development of a Li(alloy)/iron sulfide battery system for energy storage, are briefly described as an example of the way that such an understanding is being sought by coordinated interdisciplinary research. 3 figures.

  3. The advanced test reactor national scientific user facility: advancing nuclear technology education

    International Nuclear Information System (INIS)

    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 designated the Idaho National Laboratory (INL) Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The ATR NSUF provides education programs including a Users Week, internships, faculty student team projects and faculty/staff exchanges. In addition, the ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. (author)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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 1995-1999. It was included into the IAEA's Programme following endorsement in 1995 by the IAEA's 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)

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

    International Nuclear Information System (INIS)

    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)

  7. Enhanced in-pile instrumentation at the advanced test reactor

    International Nuclear Information System (INIS)

    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)

  8. Enhanced In-Pile Instrumentation at the Advanced Test Reactor

    International Nuclear Information System (INIS)

    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 (INL). To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility (NSUF) 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 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.

  9. Mirror Advanced Reactor Study engineering overview

    International Nuclear Information System (INIS)

    The Mirror Advanced Reactor Study (MARS) was the first comprehensive conceptual design of a commercial tandem mirror reactor with thermal barriers. The design exploited the inherent attractive features of a tandem mirror: steady state operation, linear central cell, simple high performance blankets, low first wall heat fluxes, natural impurity diversion by the halo plasma, no driven plasma currents or associated disruptions, and direct conversion of the charged particle power lost out the ends. The study introduced new design concepts in high field magnets, neutral beams, ECRH systems, drift pumping, direct conversion, lithium-lead blankets and plant safety. The MARS design would produce 1200 MWsub(e) net and more than 1500 MWsub(e) gross from only 2600 MW of fusion power. This high efficiency is achieved through a combination of blanket design and direct conversion. Special emphasis was placed on fusion's potential for inherent safety, lower activation and simpler disposal of radioactive waste as compared with fission. The blanket has a very low tritium inventory, cannot melt in loss-of-coolant and/or loss-of-flow accidents and can be disposed of as low level waste subject to near-surface burial. MARS would produce busbar electricity at about 7 cents per kilowatthour (constant 1983 dollars). This value is near the upper end of the cost range for new generation capability being installed in the late 1980's. Significant cost reductions can be gained by further improvements in the engineering designs combined with a simplified end cell. The largest cost reductions from engineering can be attained through redesigned magnets, heat transport system and electrical system. The combination of engineering and physics improvements are projected to lower the cost of electricity by about 40% without sacrificing the environmental, safety and maintainability attributes of MARS. This work is now being pursued in the MINIMARS study. (orig.)

  10. Heavy water moderated reactors advances and challenges

    International Nuclear Information System (INIS)

    Nuclear energy is now considered a key contributor to world electricity production, with total installed capacity nearly equal to that of hydraulic power. Nevertheless, many important challenges lie ahead. Paramount among these is gaining public acceptance: this paper makes the basic assumption that public acceptance will improve if, and only if, nuclear power plants are operated safely and economically over an extended period of time. The first task, therefore, is to ensure that these prerequisites to public acceptance are met. Other issues relate to the many aspects of economics associated with nuclear power, include capital cost, operation cost, plant performance and the risk to the owner's investment. Financing is a further challenge to the expansion of nuclear power. While the ability to finance a project is strongly dependent on meeting public acceptance and economic challenges, substantial localisation of design and manufacture is often essential to acceptance by the purchaser. The neutron efficient heavy water moderated CANDU with its unique tube reactor is considered to be particularly well qualified to respond to these market challenges. Enhanced safety can be achieved through simplification of safety systems, design of the moderator and shield water systems to mitigate severe accident events, and the increased use of passive systems. Economics are improved through reduction in both capital and operating costs, achieved through the application of state-of-the-art technologies and economy of scale. Modular features of the design enhance the potential for local manufacture. Advanced fuel cycles offer reduction in both capital costs and fuelling costs. These cycles, including slightly enriched uranium and low grade fuels from reprocessing plants can serve to increase reactor output, reduce fuelling cost and reduce waste production, while extending resource utilisation. 1 ref., 1 tab

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

    International Nuclear Information System (INIS)

    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

  12. A multipurpose research reactor design using MCNP

    International Nuclear Information System (INIS)

    The Australian Replacement Research Reactor (RRR) is one of the most recently built advanced neutron research facilities. It is a 20 megawatt open-pool reactor fueled with low enriched uranium and cooled by forced light water. The core is located inside a chimney, surrounded by heavy water as reflector. This paper describes modeling and simulation of the RRR using MCNP. Three changes in the core design are also suggested and simulated. Neutron flux distribution and k(eff) for each model is calculated and compared with those of the original model. Model A is the original RRR design. It is modeled as close as possible to the original design for benchmark and comparison purposes. In the second model (Model B), a vertical square cavity is added in the center of the core, thus providing an irradiation channel with high harder-spectrum neutron flux. A simulation shows that a fast flux as high as 3.0*1014 n/cm2.s is available in a cavity whose area is 64 cm2 while minimally disturbing the rest of the core. The original central cross-shaped control blade is split into four smaller pieces and moved to outer regions. In the third model (Model C), control blades are placed asymmetrically, leading to higher thermal flux in some locations in the reflector, which can be used, for example, for cold neutron source. In the last model (Model D), the control blades never occupy the central part of the core leading to a flux trap and higher harder-spectrum flux around z-equals-0 plane in the central cavity. Individual or combination of these changes may be incorporated in future research reactor designs

  13. Advances in Process Intensification through Multifunctional Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    O' Hern, Timothy [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Engineering Sciences Center; Evans, Lindsay [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Materials Sciences and Engineering Center; Miller, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Materials Sciences and Engineering Center; Cooper, Marcia [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Energetic Components Realization Center; Torczynski, John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pena, Donovan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gill, Walt [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Engineering Sciences Center

    2011-02-01

    This project was designed to advance the art of process intensification leading to a new generation of multifunctional chemical reactors utilizing pulse flow. Experimental testing was performed in order to fully characterize the hydrodynamic operating regimes associated with pulse flow for implementation in commercial applications. Sandia National Laboratories (SNL) operated a pilot-scale multifunctional reactor experiment for operation with and investigation of pulse flow operation. Validation-quality data sets of the fluid dynamics, heat and mass transfer, and chemical kinetics were acquired and shared with Chemical Research and Licensing (CR&L). Experiments in a two-phase air-water system examined the effects of bead diameter in the packing, and viscosity. Pressure signals were used to detect pulsing. Three-phase experiments used immiscible organic and aqueous liquids, and air or nitrogen as the gas phase. Hydrodynamic studies of flow regimes and holdup were performed for different types of packing, and mass transfer measurements were performed for a woven packing. These studies substantiated the improvements in mass transfer anticipated for pulse flow in multifunctional reactors for the acid-catalyzed C4 paraffin/olefin alkylation process. CR&L developed packings for this alkylation process, utilizing their alkylation process pilot facilities in Pasadena, TX. These packings were evaluated in the pilot-scale multifunctional reactor experiments established by Sandia to develop a more fundamental understanding of their role in process intensification. Lummus utilized the alkylation technology developed by CR&L to design and optimize the full commercial process utilizing multifunctional reactors containing the packings developed by CR&L and evaluated by Sandia. This hydrodynamic information has been developed for multifunctional chemical reactors utilizing pulse flow, for the acid-catalyzed C4 paraffin/olefin alkylation process, and is now accessible for use in

  14. Research advances in geriatric depression

    OpenAIRE

    Alexopoulos, George S.; KELLY JR., ROBERT E.

    2009-01-01

    Technical advances have facilitated the exploration of factors related to geriatric depression and have helped generate novel biological and psychosocial treatment approaches. This review summarizes the main advancements in epidemiology, clinical presentation and course, genetics, and other areas of biological research. Treatment interventions outlined in this paper include electroconvulsive therapy, repetitive transcranial magnetic stimulation, magnetic seizure therapy, vagus nerve stimulati...

  15. The search for advanced remote technology in fast reactor reprocessing

    International Nuclear Information System (INIS)

    Research and development in fast reactor reprocessing has been under way ∼ 20 yr in several countries. During the past decade, France and the United Kingdom have developed active programs in breeder reprocessing. Actual fuels from their demonstration reactors have been reprocessed in small-scale facilities. Early US work in breeder reprocessing was carried out at the Experimental Breeder Reactor II (EBR-II) facilities with the early metal fuels, and interest has renewed recently in metal fuels. A major, comprehensive program, focused on oxide fuels, has been carried out in the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) since 1974. The Federal Republic of Germany (FRG) and Japan have also carried out development programs in breeder reprocessing, and Japan appears committed to major demonstration of breeder reactors and their fuel cycles. While much of the effort in these programs addressed process chemistry and process hardware, a significant element of many of these programs, particularly the CFRP, has been on advancements in facility concepts and remote maintenance features. This paper focuses on the search for improved facility concepts and better maintenance systems in the CFRP, and, in turn, on how developments at ORNL have influenced the technology elsewhere

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

    International Nuclear Information System (INIS)

    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)

  17. IAEA's Cross Cutting Activities on Research Reactors

    International Nuclear Information System (INIS)

    Full text: For nuclear research and technology development to continue to advance, research reactors (RRs) must be safely and reliably operated, adequately utilized, refurbished when necessary, provided with adequate proliferation-resistant fuel cycle services and safely decommissioned at the end of life. The IAEA has established its competence in the area of RRs with a long history of assistance to Member States in improving their utilization, by taking the lead in the development of safety standards, norms and dissemination of information on good practices for all aspects of the nuclear fuel cycle and in the planning and implementation of decommissioning. IAEA activities on RRs are formulated to cover a broad range of RR issues and to promote the continued development of scientific research and technological development using RRs. Member States look to the IAEA for coordination of the worldwide effort in this area and for help in solving specific problems. Today RR operating organizations need to overcome challenges such as the on-going management of ageing facilities, pressures for increased vigilance with respect to non-proliferation, and shrinking resources (financial as well as human) while fulfilling an expanding role in support of nuclear technology development. The IAEA coordinates and implements an array of activities that together provide broad support for RRs. As with other aspects of nuclear technology, RR activities within the IAEA are spread through diverse groups in different Departments. To ensure harmonized approaches a Cross-cutting coordination Group on Research Reactors (CCCGRR) has been established, with representatives from all IAEA Departments actively supporting RR activities. Utilization and application activities are generally lead from within the Department of Nuclear Sciences and Applications (NA). With respect to RRs, NA is primarily carrying out IAEA activities to assist and advise Member States in assessing their needs for research

  18. Profile: Returning Argonne researcher aims to push computing speed to next level

    CERN Multimedia

    Merrion, P

    2002-01-01

    Paul Messina, one of the world's leading supercomputer scientists has returned to Argonne National Laboratory. He will split time between Argonne, where he holds the title of senior computer scientist, and Geneva, Switzerland, where he is an adviser to the director of CERN (1 page).

  19. Research nuclear reactor RA - Annual Report 2000

    International Nuclear Information System (INIS)

    Activities related to revitalisation of the RA reactor started in 1986 were fulfilled except the exchange of the complete reactor instrumentation. Since 1992, due to economic and political reasons, RA reactor is in a difficult situation. The old RA reactor instrumentation was dismantled. Decision about the future status of the reactor should be made because the aging of all the components is becoming dramatic. Control and maintenance of the reactor components was done regularly and efficiently. The most important activity and investment in 1998 was improvement of conditions for spent fuel storage in the existing pools at the RA reactor. Russian company ENTEK and IAEA are involved in this activity which was initiated 1997. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. Research reactor RA Annual report for year 2000 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection

  20. Research nuclear reactor RA - Annual Report 1998

    International Nuclear Information System (INIS)

    Activities related to revitalisation of the RA reactor started in 1986 were fulfilled except the exchange of the complete reactor instrumentation. Since 1992, due to economic and political reasons, RA reactor is in a difficult situation. The old RA reactor instrumentation was dismantled. Decision about the future status of the reactor should be made because the aging of all the components is becoming dramatic. Control and maintenance of the reactor components was done regularly and efficiently. The most important activity and investment in 1998 was improvement of conditions for spent fuel storage in the existing pools at the RA reactor. Russian company ENTEK and IAEA are involved in this activity which was initiated 1997. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. Research reactor RA Annual report for year 1998 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection

  1. Research nuclear reactor RA - Annual Report 1999

    International Nuclear Information System (INIS)

    Activities related to revitalisation of the RA reactor started in 1986 were fulfilled except the exchange of the complete reactor instrumentation. Since 1992, due to economic and political reasons, RA reactor is in a difficult situation. The old RA reactor instrumentation was dismantled. Decision about the future status of the reactor should be made because the aging of all the components is becoming dramatic. Control and maintenance of the reactor components was done regularly and efficiently. The most important activity and investment in 1998 was improvement of conditions for spent fuel storage in the existing pools at the RA reactor. Russian company ENTEK and IAEA are involved in this activity which was initiated 1997. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. Research reactor RA Annual report for year 1998 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection

  2. Usage of burnable poison on research reactors

    International Nuclear Information System (INIS)

    The fuel assemblies with burnable poison are widely used on power reactors, but there are not commonly used on research reactors. This paper shows a neutronic analysis of the advantages and disadvantages of the burnable poison usage on research reactors. This paper analyses both burnable poison design used on research reactors: Boron on the lateral wall and Cadmium wires. Both designs include a parametric study on the design parameters like the amount and geometry of the burnable poison. This paper presents the design flexibility using burnable poisons, it does not find an optimal or final design, which it will strongly depend on the core characteristics and fuel management strategy. (author)

  3. The Canadian research reactor spent fuel situation

    International Nuclear Information System (INIS)

    This paper summarizes the present research reactor spent fuel situation in Canada. The research reactors currently operating are listed along with the types of fuel that they utilize. Other shut down research reactors contributing to the storage volume are included for completeness. The spent fuel storage facilities associated with these reactors and the methods used to determine criticality safety are described. Finally the current inventory of spent fuel and where it is stored is presented along with concerns for future storage. (author). 3 figs

  4. The Berkeley TRIGA Mark III research reactor

    International Nuclear Information System (INIS)

    The Berkeley Research Reactor went critical on August 10, 1966, and achieved licensed operating power of 1000 kW shortly thereafter. Since then, the reactor has operated, by and large, trouble free on a one-shift basis. The major use of the reactor is in service irradiations, and many scientific programs are accommodated, both on and off campus. The principal off-campus user is the Lawrence Radiation Laboratory at Berkeley. The reactor is also an important instructional tool in the Nuclear Engineering Department reactor experiments laboratory course, and as a source of radioisotopes for two other laboratory courses given by the Department. Finally, the reactor is used in several research programs conducted within the Department, involving studies with neutron beams and in reactor kinetics

  5. 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)

  6. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  7. United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support

    Energy Technology Data Exchange (ETDEWEB)

    Douglas Morrell

    2011-03-01

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  8. Research reactor and its application in Thailand

    International Nuclear Information System (INIS)

    The first Thai Research Reactor (TRR-1) was established in 1961. TRR-1 had been operated with power of 1 MW from 1962 to 1975 and was shut down for modification during 1975 to 1977. The Thai Research Reactor1/modification1 (TRR-1/M1) is a multipurpose reactor with nominal power of 2 MW. Since 1977 TRR-1/M1 has been operated and utilized for various applications such as neutron activation analysis, radioisotope production, gem irradiation, neutron radiography and research works. To expand and promote the utilization of research reactor, the new 10 MW Research Reactor will be established in the Ongkarak Nuclear Research Center (ONRC) project and the project will be finished in the near future. (author)

  9. IAEA programme on research reactor safety

    International Nuclear Information System (INIS)

    This paper describes the IAEA programme on research reactor safety and includes the safety related areas of conversions to the use of low enriched uranium (LEU) fuel. The program is based on the IAEA statutory responsibilities as they apply to the requirements of over 320 research reactors operating around the world. The programme covers four major areas: (a) the development of safety documents; (b) safety missions to research reactor facilities; (c) support of research programmes on research reactor safety; (d) support of Technical Cooperation projects on research reactor safety issues. The demand for these activities by the IAEA member states has increased substantially in recent years especially in developing countries with increasing emphasis being placed on LEU conversion matters. In response to this demand, the IAEA has undertaken an extensive programme for each of the four areas above. (author)

  10. Advanced Test Reactor National Scientific User Facility Progress

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation 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. Goals of the ATR NSUF are to define the cutting edge of nuclear technology research in high temperature and radiation environments, contribute to improved industry performance of current and future light water reactors, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives

  11. Overview of research reactor operation within AECL

    International Nuclear Information System (INIS)

    This paper presents information on reactor operations within the Research Company of Atomic Energy of Canada (AECL) today relative to a few years ago, and speculates on future operations. In recent years, the need for Research Company reactors has diminished. This, combined with economic pressures, has led to the shutdown of some of the company's major reactors. However, compliance with the government agenda to privatize government companies in Canada, and a Research Company policy of business development, has led to some offsetting activities. The building of a pool-type 10 MWt MAPLE (Multipurpose Applied Physics Lattice Experimental) reactor for isotope production will assist in the sale of the AECL isotopes marketing company. A Low Enriched Uranium (LEU) fuel fabrication facility and a Tritium Extraction Plant (TEP), both currently under construction, are needed in support of the NRU (National Research Universal) reactor and are in line with business development strategies. The research program demands on NRU stretch many years into the future and the strategies for achieving effective operation of this aging reactor, now 32 years old, are discussed. The repair of the leaking light-water reflector of the NRU reactor is highlighted. The isotope business requires that a second reactor be available for back-up production and the operation of the 42 year old NRX (National Research Experimental) reactor in its present 'hot standby' mode is believed to be unique in the world

  12. Fuel Management at the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pham, V.L.; Nguyen, N.D.; Luong, B.V.; Le, V.V.; Huynh, T.N.; Nguyen, K.C. [Nuclear Research Institute, 01 Nguyen Tu Luc Street, Dalat City (Viet Nam)

    2011-07-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the old 250 kW TRIGA-MARK II reactor. The spent fuel storage was newly designed and installed in the place of the old thermalizing column for biological irradiation. The core was loaded by Russian WWR-M2 fuel assemblies (FAs) with 36% enrichment. The reconstructed reactor reached its initial criticality in November 1983 and attained it nominal power of 500 kW in February 1984. The first fuel reloading was executed in April 1994 after more than 10 years of operation with 89 highly enriched uranium (HEU) FAs. The third fuel reloading by shuffling of HEU FAs was executed in June 2004. After the shuffling the working configuration of reactor core kept unchanged of 104 HEU FAs. The fourth fuel reloading was executed in November 2006. The 2 new HEU FAs were loaded in the core periphery, at previous locations of wet irradiation channel and dry irradiation channel. After reloading the working configuration of reactor core consisted of 106 HEU FAs. Contracts for reactor core conversion between USA, Russia, Vietnam and the International Atomic Energy Agency for Nuclear fuel manufacture and supply for DNRR and Return of Russian-origin non-irradiated highly enriched uranium fuel to the Russian Federation have been realized in 2007. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory and Vietnam Atomic Energy Institute the mixed core configurations of irradiated HEU and new low enriched uranium (LEU) FAs has been created on 12 September, 2007 and on 20 July, 2009. After reloading in 2009, the 14 HEU FAs with highest burnup were removed from the core and put in the interim storage in reactor pool. The works on full core conversion for the DNRR are being realized in cooperation with the organizations, DOE and IAEA. Contract for Nuclear fuel manufacture and supply of 66 LEU FAs for DNRR

  13. Utilisation of British University Research Reactors.

    Science.gov (United States)

    Duncton, P. J.; And Others

    British experience relating to the employment of university research reactors and subcritical assemblies in the education of nuclear scientists and technologists, in the training of reactor operators and for fundamental pure and applied research in this field is reviewed. The facilities available in a number of British universities and the uses…

  14. Problems of Decommissioning Research Reactor IR-100

    International Nuclear Information System (INIS)

    The research reactor IR-100 with a thermal power of 200 kW is assigned to conduct science research and training activities in the fields of nuclear and molecular physics, radiation chemistry, radioactive isotope production, material, irradiation in neutron and gamma fields of devices and equipment, as well as for training of specialists for nuclear reactor operation

  15. The advanced light water reactor program approach for gaining acceptance

    International Nuclear Information System (INIS)

    Electric utilities in the US face several obstacles and disincentives to building new nuclear power plants. A reformed licensing process that permits true one-step licensing is a prerequisite, as is timely implementation of the technical and political solutions of the high-level waste issue by the federal government and low-level waste storage by state governments. In addition, a more tangible acceptance of the need, benefits, and residual risk of nuclear power by all concerned institutions is an essential impetus for the return of the nuclear option. An improved version of the light water reactor (LWR) is expected to be the preferred choice for the next increment of nuclear capacity ordered in the US. The paper discusses the need for research and advanced LWR development. The advanced LWR program, sponsored jointly by the utility industry and the US Department of Energy (DOE), offers the most promising approach to developing the next generation of nuclear power plants

  16. The future role of research reactors

    International Nuclear Information System (INIS)

    The decline of neutron source capacity in the next decades urges for the planning and construction of new neutron sources for basic and applied research with neutrons. Modern safety precautions of research reactors make them competitive with other ways of neutron production using non-chain reactions for many applications. Research reactors consequently optimized offer a very broad range of possible applications in basic and applied research. Research reactors at universities also in the future have to play an important role in education and training in basic and applied nuclear science. (orig.)

  17. Developments in the regulation of research reactors

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has data on over 670 research reactors in the world. Fewer than half of them are operational and a significant number are in a shutdown but not decommissioned state. The International Nuclear Safety Advisory Group (INSAG) has expressed concerns about the safety of many research reactors and this has resulted in a process to draw up an international Code of Conduct on the Safety of Research Reactors. The IAEA is also reviewing its safety standards applying to research reactors. On the home front, regulation of the construction of the Replacement Research Reactor continues. During the construction phase, regulation has centred around the consideration of Requests for Approval (RFA) for the manufacture and installation of systems, structures and components important for safety. Quality control of construction of systems, structures and components is the central issue. The process for regulation of commissioning is under consideration

  18. Nuclear data usage for research reactors

    International Nuclear Information System (INIS)

    In the department of research reactor, many neutronics calculations have been performed to construct, to operate and to modify research reactors of JAERI with several kinds of nuclear data libraries. This paper presents latest two neutronic analyses on research reactors. First one is design work of a low enriched uranium (LEU) fuel for JRR-4 (Japan Research Reactor No.4). The other is design of a uranium silicon dispersion type (silicide) fuel of JRR-3M (Japan Research Reactor No.3 Modified). Before starting the design work, to estimate the accuracy of computer code and calculation method, experimental data are calculated with several nuclear data libraries. From both cases of calculations, it is confirmed that JENDL-3.2 gives about 1 %Δk/k higher excess reactivity than JENDL-3.1. (author)

  19. Advanced Remote Sensing Research

    Science.gov (United States)

    Slonecker, Terrence; Jones, John W.; Price, Susan D.; Hogan, Dianna

    2008-01-01

    'Remote sensing' is a generic term for monitoring techniques that collect information without being in physical contact with the object of study. Overhead imagery from aircraft and satellite sensors provides the most common form of remotely sensed data and records the interaction of electromagnetic energy (usually visible light) with matter, such as the Earth's surface. Remotely sensed data are fundamental to geographic science. The Eastern Geographic Science Center (EGSC) of the U.S. Geological Survey (USGS) is currently conducting and promoting the research and development of three different aspects of remote sensing science: spectral analysis, automated orthorectification of historical imagery, and long wave infrared (LWIR) polarimetric imagery (PI).

  20. Jordan Research and Training Reactor (JRTR) Utilization Facilities

    International Nuclear Information System (INIS)

    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 (U3Si2) 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 1.45×1014 cm-2s-1, and is controlled by a Hafnium control absorber rod and B4C 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)

  1. Research reactor job analysis - A project description

    International Nuclear Information System (INIS)

    Addressing the need of the improved training in nuclear industry, nuclear utilities established training program guidelines based on Performance-Based Training (PBT) concepts. The comparison of commercial nuclear power facilities with research and test reactors owned by the U.S. Department of Energy (DOE), made in an independent review of personnel selection, training, and qualification requirements for DOE-owned reactors pointed out that the complexity of the most critical tasks in research reactors is less than that in power reactors. The U.S. Department of Energy (DOE) started a project by commissioning Oak Ridge Associated Universities (ORAU) to conduct a job analysis survey of representative research reactor facilities. The output of the project consists of two publications: Volume 1 - Research Reactor Job Analysis: Overview, which contains an Introduction, Project Description, Project Methodology,, and. An Overview of Performance-Based Training (PBT); and Volume 2 - Research Reactor Job Analysis: Implementation, which contains Guidelines for Application of Preliminary Task Lists and Preliminary Task Lists for Reactor Operators and Supervisory Reactor Operators

  2. Development of Digital MMIS for Research Reactors: Graded Approaches

    International Nuclear Information System (INIS)

    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

  3. Software development for research reactors

    International Nuclear Information System (INIS)

    The Texas A and M University Nuclear Science Center, in a program jointly sponsored with the International Atomic Energy Agency, is developing a series of computer software programs of use at research reactor facilities. The programs cover a wide range of topics including activation and shielding calculations, control rod calibrations, power calorimetrics, and fuel inventory including burnup. Many of the programs are modified and improved versions of programs already in use at the NSC that ran on outdated computing equipment. All of the new versions were written in Fortran77 on the NSC's new TI Pro microcomputer and are IBM-compatible. This paper describes the development and translation efforts in preparing the programs for use by other facilities, and gives an overview of the aim of the development effort. A brief description of each program that has been or is to be written is given including the required inputs and the resulting outputs. This paper also addresses the original needs that brought about the development program and the benefits to facility operations that each program provides. The programs discussed are available to interested parties in a hard-copy listing as requested. (author)

  4. The Advanced Test Reactor as a National Scientific User Facility

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) has been in operation since 1967 and mainly used to support U.S. Department of Energy (US DOE) materials and fuels research programs. Irradiation capabilities of the ATR and post-irradiation examination capabilities of the Idaho National Laboratory (INL) were generally not being utilized by universities and other potential users due largely to a prohibitive pricing structure. While materials and fuels testing programs using the ATR continue to be needed for US DOE programs such as the Advanced Fuel Cycle Initiative and Next Generation Nuclear Plant, US DOE recognized there was a national need to make these capabilities available to a broader user base. In April 2007, the U.S. Department of Energy designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). As a NSUF, most of the services associated with university experiment irradiation and post-irradiation examinations are provided free-of-charge. The US DOE is providing these services to support U.S. leadership in nuclear science, technology, and education and to encourage active university/industry/laboratory collaboration. The first full year of implementing the user facility concept was 2008 and it was a very successful year. The first university experiment pilot project was developed in collaboration with the University of Wisconsin and began irradiation in the ATR in 2008. Lessons learned from this pilot program will be applied to future NSUF projects. Five other university experiments were also competitively selected in March 2008 from the initial solicitation for proposals. The NSUF now has a continually open process where universities can submit proposals as they are ready. Plans are to invest in new and upgraded capabilities at the ATR, post-irradiation examination capabilities at the INL, and in a new experiment assembly facility to further support the implementation of the user facility concept. Through a newly created Partnership Program

  5. ADVANCED GAS TURBINE SYSTEMS RESEARCH

    Energy Technology Data Exchange (ETDEWEB)

    Unknown

    2002-04-01

    The activities of the Advanced Gas Turbine Systems Research (AGTSR) program for this reporting period are described in this quarterly report. The report is divided into discussions of Membership, Administration, Technology Transfer (Workshop/Education), Research and Miscellaneous Related Activity. Items worthy of note are presented in extended bullet format following the appropriate heading.

  6. ADVANCED GAS TURBINE SYSTEMS RESEARCH

    Energy Technology Data Exchange (ETDEWEB)

    Unknown

    2002-02-01

    The activities of the Advanced Gas Turbine Systems Research (AGTSR) program for this reporting period are described in this quarterly report. The report is divided into discussions of Membership, Administration, Technology Transfer (Workshop/Education), Research and Miscellaneous Related Activity. Items worthy of note are presented in extended bullet format following the appropriate heading.

  7. RA Research nuclear reactor - Annual report 1987

    International Nuclear Information System (INIS)

    Annual report concerning the project 'RA research nuclear reactor' for 1987, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities

  8. RA Research reactor, Annual report 1988

    International Nuclear Information System (INIS)

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities

  9. Research nuclear reactor RA - Annual Report 1989

    International Nuclear Information System (INIS)

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities

  10. Research reactor records in the INIS database

    International Nuclear Information System (INIS)

    This report presents a statistical analysis of more than 13,000 records of publications concerned with research and technology in the field of research and experimental reactors which are included in the INIS Bibliographic Database for the period from 1970 to 2001. The main objectives of this bibliometric study were: to make an inventory of research reactor related records in the INIS Database; to provide statistics and scientific indicators for the INIS users, namely science managers, researchers, engineers, operators, scientific editors and publishers, decision-makers in the field of research reactors related subjects; to extract other useful information from the INIS Bibliographic Database about articles published in research reactors research and technology. (author)

  11. Issues and future direction of thermal-hydraulics research and development in nuclear power reactors

    International Nuclear Information System (INIS)

    The paper archives the proceedings of an expert panel discussion on the issues and future direction of thermal-hydraulic research and development in nuclear power reactors held at the NURETH-14 conference in Toronto, Canada, in September 2011. Thermal-hydraulic issues related to both operating and advanced reactors are presented. Advances in thermal-hydraulics have significantly improved the performance of operating reactors. Further thermal-hydraulics research and development is continuing in both experimental and computational areas for operating reactors, reactors under construction or ready for near-term deployment, and advanced Generation-IV reactors. As the computing power increases, the fine-scale multi-physics computational models, coupled with the systems analysis code, are expected to provide answers to many challenging problems in both operating and advanced reactor designs

  12. Shielding considerations for advanced space nuclear reactor systems

    International Nuclear Information System (INIS)

    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 (UO2) 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

  13. Advanced designs of VVER reactor plant

    International Nuclear Information System (INIS)

    The history of VVER reactors, current challenges and approaches to the challenges are highlighted. The VVER-1200 reactor of 3+ generation for AES-2006 units are under construction at the Leningrad 2 nuclear power plant (LNPP-2). The main parameters are listed and details are presented of the vessel, steam generator, and improved fuel. The issue of the NPP safety is discussed. Additional topics include the MIR-1200 reactor unit, VVER-600, and VVER-SCP (Generation 4). (P.A.)

  14. 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

  15. Manual for the operation of research reactors

    International Nuclear Information System (INIS)

    The great majority of the research reactors in newly established centres are light-water cooled and are often also light-water moderated. Consequently, the IAEA has decided to publish in its Technical Reports Series a manual dealing with the technical and practical problems associated with the safe and efficient operation of this type of reactor. Even though this manual is limited to light-water reactors in its direct application and presents the practices and experience at one specific reactor centre, it may also be useful for other reactor types because of the general relevance of the problems discussed and the long experience upon which it is based. It has, naturally, no regulatory character but it is hoped that it will be found helpful by staff occupied in all phases of the practical operation of research reactors, and also by those responsible for planning their experimental use. 23 refs, tabs

  16. Performance of a multipurpose research electrochemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Henquin, E.R. [Programa de Electroquimica Aplicada e Ingenieria Electroquimica (PRELINE), Facultad de Ingenieria Quimica, Universidad Nacional del Litoral, Santiago del Estero 2829, S3000AOM Santa Fe (Argentina); Bisang, J.M., E-mail: jbisang@fiq.unl.edu.ar [Programa de Electroquimica Aplicada e Ingenieria Electroquimica (PRELINE), Facultad de Ingenieria Quimica, Universidad Nacional del Litoral, Santiago del Estero 2829, S3000AOM Santa Fe (Argentina)

    2011-07-01

    Highlights: > For this reactor configuration the current distribution is uniform. > For this reactor configuration with bipolar connection the leakage current is small. > The mass-transfer conditions are closely uniform along the electrode. > The fluidodynamic behaviour can be represented by the dispersion model. > This reactor represents a suitable device for laboratory trials. - Abstract: This paper reports on a multipurpose research electrochemical reactor with an innovative design feature, which is based on a filter press arrangement with inclined segmented electrodes and under a modular assembly. Under bipolar connection, the fraction of leakage current is lower than 4%, depending on the bipolar Wagner number, and the current distribution is closely uniform. When a turbulence promoter is used, the local mass-transfer coefficient shows a variation of {+-}10% with respect to its mean value. The fluidodynamics of the reactor responds to the dispersion model with a Peclet number higher than 10. It is concluded that this reactor is convenient for laboratory research.

  17. IAEA Guidelines for New Research Reactor Projects

    International Nuclear Information System (INIS)

    In recent years, the interest of the IAEA Member States in developing research reactor (RR) programmes has been steadily growing. Currently a number of Member States are in different stages of new research reactor projects. Some of these Member States are building their first research reactor as their country's introduction to nuclear science and technology infrastructure. To support Member States in such efforts, the IAEA published in 2012 a Nuclear Energy Series Report NP-T-5.1: 'Specific Considerations and Milestones for a New Research Reactor Project'. This publication provides guidance on the timely preparation of a research reactor project through a sequential development process. It includes a detailed description of the range of infrastructure issues that need to be addressed and the expected level of achievement (or milestones) at the end of each phase of the project. The publication provides a discussion of the mechanisms for justification of a research reactor, and for building stakeholder support. It includes both the technical, legal, regulatory and safety infrastructure, and the development of qualified human resources needed for a research reactor. The publication also addresses the evolution of infrastructure needs from the time a Member State first considers a research reactor and its associated facilities, through the stages of planning, bid preparation, construction, start-up, and preparation for commissioning. The subsequent stages of operation, decommissioning, spent fuel and waste management issues are addressed in the publication to the degree necessary for appropriate planning prior to research reactor commissioning. The feedback from the IAEA activities, in particular from Member States establishing their first research reactor, indicated the need for further guidance on the development of the technical specifications for the bidding process of a research reactor project. In responding to these needs, a Nuclear Energy Series Report on

  18. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  19. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  20. Education and Training on ISIS Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X. [French Atomic Energy and Alternative Energies Commission, Paris (France)

    2013-07-01

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions.

  1. Education and Training on ISIS Research Reactor

    International Nuclear Information System (INIS)

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions

  2. Reactor containment research and development

    Energy Technology Data Exchange (ETDEWEB)

    Weil, N. A.

    1963-06-15

    An outline is given of containment concepts, sources and release rates of energy, responses of containment structures, effects of projectiles, and leakage rates of radioisotopes, with particular regard to major reactor accidents. (T.F.H.)

  3. Research and development into power reactor fuel performance

    International Nuclear Information System (INIS)

    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 UO2 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

  4. Thermohydraulic and constructional boundary conditions of an advanced PWR reactor

    International Nuclear Information System (INIS)

    The advantages and special features of an advanced PWR reactor (FDWR) have been systematically investigated for several years by the Department of Space Flight and Reactor Technology of the University of Brunswick (LRR-TUBS). The FDWR will have a homogeneous core, i.e. the fuel elements will consist of fuel rods of the same size and enrichment. (orig./GL)

  5. Economics of an advanced pressurized water reactor (APWR)

    International Nuclear Information System (INIS)

    Limited natural uranium resources together with their low utilization in current lightwater reactors (LWR) on the one hand and the high capital investments for a LWR and fast breeder reactor system resulting in a high fuel utilization are the most important reasons for research and development (R+D) work related to a high converting APWR system. It is the main task of this analysis to study the economic behaviour of an APWR combining its technical and physical parameters with an economic data base. After introductional remarks chapter II presents the potential improvements of the uranium utilization and their importance for a whole national economy. Chapter III shows the micro-economic aspects for an electricity producing utility using such an APWR plant. The chapter is restricted to the fuel cycle costs and their dependence on various parameters. The corresponding costs of other nuclear power plants are described in chapter IV and compared to those of the APWR in chapter V. Finally a cost comparison on the basis of the electricity generating costs will complete the economic picture of an advanced pressurized water reactor. (orig./UA)

  6. Progress in Promoting Research Reactor Coalitions

    International Nuclear Information System (INIS)

    This presentation treats of the IAEA's role in Promoting Research Reactor (RR) coalitions, presents the strategic view, the types of coalitions, the 2007-2008 activities and Results, and the upcoming activities. The RR Coalitions Progress is presented first (Initial discussions, project design, approval of NTI grant request, informal consultations and development of 'national' proposals, Number of 'models' identified, exploratory missions/meetings, initial implementation of several coalitions, IAEA coordination, ideas/proposals/ventures, initial support. Some countries, institutes, or users want access to reactor capabilities without, or in advance of, building a domestic facility. Some countries, institutes, or users need access to alternative capabilities to permit the closure/consolidation of marginal facilities. Cooperative arrangements will result in increased utilization for each participant. The results from the reactor view are as follows: cover increases in order levels or scientific research; cover facility outages (planned or un-planned); delegate 'less profitable' products and services; access capacity for new products and services; reduce transport needs by geographical optimization; reduce investment needs by contracting for complementary capabilities; reduce costs of medical radio-isotope for R and D; share best practices in operations and safety. The results from the stakeholder View are: Better information on what reactors can offer/provide; greater range of services; more proactive product and service support; greater reliability in supplies of products and services. The types of coalitions are of different forms to meet needs, capabilities, objectives of members. In general they start small, evolve, change form, expand as confidence grows. The role of the Scientific consortium is to: distribute excess demand, test new concepts for implementation at high-flux reactors, direct requests for access to most appropriate RR, share best practices

  7. Advanced tokamak concepts and reactor designs

    NARCIS (Netherlands)

    Oomens, A. A. M.

    2000-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described, some examples

  8. Impact of proposed research reactor standards on reactor operation

    International Nuclear Information System (INIS)

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  9. Research reactor utilization in chemistry programmes

    International Nuclear Information System (INIS)

    The establishment and roles of the Philippines Atomic Energy Commission in promoting and regulating the use of atomic energy are explained. The research reactor, PRR-1 is being converted to TRIGA to meet the increasing demands of high-flux. The activities of PAEC in chemistry research programs utilizing reactor are discussed in detail. The current and future plans of Research and Development programs are also included. (A.J.)

  10. Operational experience with research reactors in Trombay

    International Nuclear Information System (INIS)

    The research reactors Apsara and Cirus, located at the Bhabha Atomic Research Centre, Trombay, Bombay have recently completed 30 years and 26 years of successful operation respectively. Both reactors have been used extensively for research, isotope production and manpower training. Several measures have been taken towards achieving this long successful operation. These include preventive maintenance, meticulous control on chemistry of fluid systems, proper physics management, surveillance measures and modifications to system and equipment based on experience. Extensive training of O and M personnel has been another important factor contributing towards this. Major topics detailed in this paper include detection of one leaky reactor vessel lattice tube in Cirus and its successful plugging, replacement of old control system hardware consisting of vacuum tubes with modern solid state devices in both the reactors, installation of S.S.liner in Apsara reactor pool and certain modifications towards improving fuel performance in Cirus. Some aspects of preventive maintenance are also dealt with. (author)

  11. Recent Research Progress on Non-aqueous Lithium-Air Batteries from Argonne National Laboratory

    Directory of Open Access Journals (Sweden)

    Jun Lu

    2013-11-01

    Full Text Available Rechargeable non-aqueous Li-air battery technology offers potential advantages over other existing battery systems in terms of specific energy and energy density, which could enable the driving range of an electric vehicle to be comparable to that of gasoline vehicles. Development of efficient cathode catalysts and stable electrolytes for the Li-air battery has been intensively investigated for the past several years, and a number of review articles covering different topics are already available. This review mainly focuses on the research activities on rechargeable non-aqueous Li-air batteries at Argonne National Laboratory, with the emphasis on the gains in understanding of electrolyte decomposition, the structure and magnetic properties of lithium peroxide (Li2O2, development of an air-breathing cathode, and the effect of oxygen crossover on the lithium anode. Insights from this research have led to the improvement of the electrochemical performance of Li-air batteries. Promising paths for future work on rechargeable Li-air batteries are also discussed.

  12. Nuclear fuel cycle programs of Argonne's Chemical Engineering Division

    International Nuclear Information System (INIS)

    Argonne National Laboratory's Chemical Engineering Division is actively involved in the research, development and demonstration of nuclear fuel cycle technologies for the United States Department of Energy Advanced Fuel Cycle Initiative, Generation IV, and Yucca Mountain programs. This paper summarizes current technology development initiatives within the Division that address the needs of the United States' advanced nuclear energy programs. (authors)

  13. Evaluation of the trial studies for a advanced marine reactor, (8)

    International Nuclear Information System (INIS)

    JAERI carried out the trial design of three type reactors (semi-integrated type, integrated type, and self pressurized integrated type) in order to clarity reactor type of advanced marine reactor for the best practical use in future and to extract its research and development theme. This report presents the comparison and estimation of these three type reactors from economical view point. Economical comparison carried out with next items for estimating the operating cost of ships powered by these three type reactors. (1) cost of reactor plants. (2) cost of fuel processing. (3) fuel cycle cost. As a result, it is cleared that the operating cost is the lowest in the ship powered by the integrated type reactor. (author)

  14. Safety of Ghana Research Reactor (GHARR-1)

    International Nuclear Information System (INIS)

    The Ghana Research Reactor, GHARR-1 is a low power research rector with maximum thermal power lever of 30kW. The reactor is inherently safe and uses highly enriched uranium (HEU) as fuel, light water as moderator and beryllium as a reflector. The construction, commissioning and operation of this reactor have been subjected to the system of authorization and inspection developed by the Regulatory Authority, the Radiation Protection Board (RPB) with the assistance of the International Atomic Energy Agency. The reactor has been regulated by the preparation of an Interim Safety Analysis Report (SAR) based upon International Atomic Energy Agency standards. An International Safety Assessment peer review and safe inspections have confirmed a high level of operational safety of the reactor since it started operation in 1994. Since its operation there has been no significant reported incident/accidents. Several studies have validated the inherent safety of the reactor. The reactor has been used for neutron activation analysis of various samples, research and teaching. About 1000 samples are analysed annually. The final Safety Analysis Report (SAR) was submitted (after five years of extensive research on the operational reactor) to the Regulatory Authority for review in June 2000. (author)

  15. Utilization of the SLOWPOKE-2 research reactor

    International Nuclear Information System (INIS)

    SLOWPOKEs are typically low power research reactors that have a limited number of applications. However, a significant range of NAA can be performed with such reactors. This paper describes a SLOWPOKE-based NAA program that is performing a valuable series of studies in Jamaica, including geological mapping and pollution assessment. (author)

  16. No small fry: Decommissioning research reactors

    International Nuclear Information System (INIS)

    To get a permit to build a research reactor, would-be operators need to submit an initial decommissioning plan for the eventual shutdown of their new facility. This, however, was not a requirement back in the 1950s, 60s and 70s when most research reactors that are now nearing the end of their working lives were built. The result: many unused reactors sit idle in the middle of university campuses, research parks and hospital compounds, because their operators lack the proper plans to decommission them

  17. Effective utilization and management of research reactors

    International Nuclear Information System (INIS)

    The problem of utilizing a research reactor effectively is closely related to its management and therefore should not be considered separately. Too often, attention has been focused on specific techniques and methods rather than on the overall programme of utilization, with the result that skills and equipment have been acquired without any active continuing programme of applications and services. The seminar reported here provided a forum for reactor managers, users, and operators to discuss their experience. At the invitation of the Government of Malaysia, it was held at the Asia Pacific Development Centre, Kuala Lumpur, from 7 to 11 November 1983. It was attended by about 50 participants from 19 Member States; it is hoped that a report on the seminar, including papers presented, can be published and thus reach a wider audience. Thirty-one lectures and contributions were presented at a total of seven sessions: Research reactor management; Radiation exposure and safety; Research reactor utilization (two sessions); PUSPATI Research Reactor Project Development; Core conversion to low-enriched uranium, and safeguards; Research reactor technology. In addition, a panel discussed the causes and resolutions of the under-utilization of research reactors

  18. The first university research reactor in India

    International Nuclear Information System (INIS)

    As the first university research reactor in India, the low power, pool type with fixed core and low enriched uranium fuel research reactor is under construction in the Andhra university campus, Andhra Pradesh, India. The reactor is expected to be commissioned during 2001-2002. The mission of the reactor is to play the research center as a regional research facility catering to the needs of academic institutions and industrial organizations of this region of the country. Further, to encourage interdisplinary and multidisplinary research activities, to supply radioisotope and labelled compounds to the user institutions and to create awareness towards the peaceful uses of atomic energy. This report describes its objectives, status and future plans in brief. (H. Itami)

  19. 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.

  20. RESEARCH ADVANCES IN ANAEROBIC BAFFLED REACTOR (ABR)%折流式厌氧反应器(ABR)的研究进展

    Institute of Scientific and Technical Information of China (English)

    王建龙; 韩英健; 钱易

    2000-01-01

    Anaerobic baffled reactor (ABR) is a new high efficient anaerobic treatmen t system. ABR has some advantages over other anaerobic reactors, such as sim ple design,inexpensive to construct,stable to hydraulic and organic shock load ings, and high treatment efficiency. Different types of ABR are described, the study progress and application situation are summarized, and the future development is prospected in this paper. Fig 2, Tab 2, Ref 45