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Sample records for advanced reactor study

  1. 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

  2. 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.

  3. 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

  4. 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)

  5. 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.)

  6. 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

  7. 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

  8. Methods for studying fuel management in advanced gas cooled reactors

    International Nuclear Information System (INIS)

    The methods used for studying fuel and absorber management problems in AGRs are described. The basis of the method is the use of ARGOSY lattice data in reactor calculations performed at successive time steps. These reactor calculations may be quite crude but for advanced design calculations a detailed channel-by-channel representation of the whole core is required. The main emphasis of the paper is in describing such an advanced approach - the ODYSSEUS-6 code. This code evaluates reactor power distributions as a function of time and uses the information to select refuelling moves and determine controller positions. (author)

  9. Mirror Advanced Reactor Study (MARS) final report summary

    International Nuclear Information System (INIS)

    The Mirror Advanced Reactor Study (MARS) has resulted in an overview of a first-generation tandem mirror reactor. The central cell fusion plasma is self-sustained by alpha heating (ignition), while electron-cyclotron resonance heating and negative ion beams maintain the electrostatic confining potentials in the end plugs. Plug injection power is reduced by the use of high-field choke coils and thermal barriers, concepts to be tested in the Tandem Mirror Experiment-Upgrade (TMX-U) and Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory

  10. Studies of a modular advanced stellarator reactor ASRA6C

    International Nuclear Information System (INIS)

    This study is directed towards the clarification of critical issues of advanced modular stellerator reactors exploiting the inherent potential of steady state operation, and is not a point design study of a reactor. Critical technology issues arise from the three-dimensional magnetic field structure. The first wall, blanket and shield are more complex than those of axi-symmetric systems, but this is eased at moderate to large aspect ratio typical of stellarators. Several blanket options have been studied and a thin blanket (21 cm) was the first choice for the design. Superconducting modular coils were investigated with respect to the conductor and mechanical supports. From the analysis of forces and stresses caused by the electromagnetic loads the coils are considered to be feasible, although shear stresses might pose a critical issue. Demountable intermagnetic support elements were designed for use at separation areas between the cryostat modules. A scheme for remote reactor maintenance was also developed. The plasma physics issues of different configurations were studied using extrapolations of transport behaviour and equilibrium from theory and present experiments. These studies indicate that the confinement and equilibrium behaviour is adequate for ignited operation at an average value of 5% beta. Impurities may pose a critical issue. Several impurity control operations were investigated; a pumped limiter configuration utilizing the 'ergodic layer' at the plasma edge was chosen for edge plasma and impurity control. A general conclusion of the study is that the modular stellerator configuration offers interesting prospects regarding the development towards steady-state reactors. (orig.)

  11. Studies of a modular advanced stellarator reactor ASRA6C

    International Nuclear Information System (INIS)

    This study is directed towards the clarification of critical issues of advanced modular stellerator reactors exploiting the inherent potential of steady state operation, and is not a point design study of a reactor. Critical technology issues arise from the three-dimensional magnetic field structure. The first wall, blanket and shield are more complex than those of axi-symmetric systems, but this is eased at moderate to large aspect ratio typical of stellerators. Several blanket options have been studied and a thin blanket (21 cm) was the first choice for the design. Superconducting modular coils were investigated with respect to the conductor and mechanical supports. From the analysis of forces and stresses caused by the electromagnetic loads the coils are considered to be feasible, although shear stresses might pose a critical issue. Demountable intermagnetic support elements were designed for use at separation areas between the cryostat modules. A scheme for remote reactor maintenance was also developed. The plasma physics issues of different configurations were studied using extrapolations of transort behaviour and equilibrium from theory and present experiments. These studies indicate that the confinement and equilibrium behaviour is adequate for ignited operation at an average value of 5% beta. Impurities may pose a critical issue. Several impurity control operations were investigated; a pumped limiter configuration utilizing the 'ergodic layer' at the plasma edge was chosen for edge plasma and impurity control. A general conclusion of the study is that the modular stellerator configuration offers interesting prospects regarding the development towards steady-state reactors. (orig.)

  12. 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.

  13. 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.

  14. Study of Pu consumption in Advanced Light Water Reactors

    International Nuclear Information System (INIS)

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

  15. Feasibility study of advanced fuel burning nuclear reactors

    International Nuclear Information System (INIS)

    An investigation has been conducted to determine both physics, engineering and economic aspects of fusion power reactors based on magnetic confinement and on burning advanced fuels (AFs). DT burning Tokamaks are taken as reference concept. We show that the attractive features of advanced fuels, in particular of neutronlean proton-based AFs, can be combined, in appropriately designed AF reactors (high beta), with power densities comparable to or even higher than those achievable in DT Tokamaks. Moreover we identify physical requirements which would assure Q values well above unity. As an example a semi-open confinement scheme is analyzed based on a self-consistent plasma calculation. We find that a mirror, even if only ''semi-open'' as a result of strong diamagnetism, can barely be expected to achieve high Q values. Therefore confinement schemes such as compact tori, multipole surmacs etc. may be required to burn AFs. We conclude that the economics of AF reactors, as determined by the nuclear boiler power density, may be superior to that of DT-rectors if low recirculating power fractions can be obtained by appropriate plasma tayloring (high fractional transfer of fusion power to ions required). A more detailed investigation is suggested for proton-based fuel cycles. (orig.)

  16. Numerical study on seismic response of the reactor coolant pump in Advanced Passive Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Highlights: • An artificial accelerogram of the specified SSE is generated. • A dynamic FE model of the RCP in AP1000 (with gyroscopic and FSI effects) is developed. • The displacement, force, moment and stress in the RCP during the earthquake are summarized. - Abstract: The reactor coolant pump in the Advanced Passive Pressurized Water Reactor is a kind of nuclear canned-motor pump. The pump is classified as Seismic Category I, which must function normally during the Safe Shutdown Earthquake. When the nuclear power plant is located in seismically active region, the seismic response of the reactor coolant pump may become very important for the safety assessment of the whole nuclear power plant. In this article, an artificial accelerogram is generated. The response spectrum of the artificial accelerogram fits well with the design acceleration spectrum of the Safe Shutdown Earthquake. By applying the finite element modeling method, the dynamic finite element models of the rotor and stator in the reactor coolant pump are created separately. The rotor and stator are coupled by the journal bearings and the annular flow between the rotor and stator. Then the whole dynamic model of the reactor coolant pump is developed. Time domain analysis which uses the improved state-space Newmark method of a direct time integration scheme is carried out to investigate the response of the reactor coolant pump under the horizontal seismic load. The results show that the reactor coolant pump responds differently in the direction of the seismic load and in the perpendicular direction. During the Safe Shutdown Earthquake, the displacement response, the shear force, the moment and the journal bearing reaction forces in the reactor coolant pump are analyzed

  17. Numerical study on seismic response of the reactor coolant pump in Advanced Passive Pressurized Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    De, Cheng, E-mail: 0100209064@sjtu.edu.cn; Zhen-Qiang, Yao, E-mail: zqyaosjtu@gmail.com; Ya-bo, Xue; Hong, Shen

    2014-10-15

    Highlights: • An artificial accelerogram of the specified SSE is generated. • A dynamic FE model of the RCP in AP1000 (with gyroscopic and FSI effects) is developed. • The displacement, force, moment and stress in the RCP during the earthquake are summarized. - Abstract: The reactor coolant pump in the Advanced Passive Pressurized Water Reactor is a kind of nuclear canned-motor pump. The pump is classified as Seismic Category I, which must function normally during the Safe Shutdown Earthquake. When the nuclear power plant is located in seismically active region, the seismic response of the reactor coolant pump may become very important for the safety assessment of the whole nuclear power plant. In this article, an artificial accelerogram is generated. The response spectrum of the artificial accelerogram fits well with the design acceleration spectrum of the Safe Shutdown Earthquake. By applying the finite element modeling method, the dynamic finite element models of the rotor and stator in the reactor coolant pump are created separately. The rotor and stator are coupled by the journal bearings and the annular flow between the rotor and stator. Then the whole dynamic model of the reactor coolant pump is developed. Time domain analysis which uses the improved state-space Newmark method of a direct time integration scheme is carried out to investigate the response of the reactor coolant pump under the horizontal seismic load. The results show that the reactor coolant pump responds differently in the direction of the seismic load and in the perpendicular direction. During the Safe Shutdown Earthquake, the displacement response, the shear force, the moment and the journal bearing reaction forces in the reactor coolant pump are analyzed.

  18. Mirror Advanced Reactor Study (MARS): executive summary and overview

    Energy Technology Data Exchange (ETDEWEB)

    Logan, B.G.; Perkins, L.J.; Gordon, J.D.

    1984-07-01

    Two self-consistent MARS configurations are discussed - a 1200-MWe commercial electricity-generating plant and a synguels-generating plant that produces hydrogen with an energy equivalent to 26,000 barrels of oil per day. The MARS machine emphasizes the attractive features of the tandem mirror concept, including steady-state operation, a small-diameter high-beta plasma, a linear central cell with simple low-maintenance blankets, low first-wall heat fluxes (<10 W/cm/sup 2/), no driven plasma currents or associated disruptions, natural halo impurity diversion, and direct conversion of end-loss charged-particle power. The MARS electric plant produces 2600 MW of fusion power in a 130-m-long central cell. Advanced tandem-mirror plasma-engineering concepts, a high-efficiency liquid lithium-lead (Li/sub 17/Pb/sub 83/) blanket, and efficient direct electrical conversion of end loss power combine to produce a high net plant efficiency of 36%. With a total capital cost of $2.9 billion (constant 1983 dollars), the MARS electric plant produces busbar electricity at approx. 7 cents/kW-hour. The MARS synfuels plant produces 3500 MW of fusion power in a 150-m-long central cell. A helium-gas-cooled silicon carbide pebble-bed blanket provides high-temperature (1000/sup 0/C) heat to a thermochemical water-splitting cycle and the resulting hydrogen is catalytically converted to methanol for distribution. With a total capital cost of $3.6 billion (constant 1983 dollars), the synfuels plant produces methanol fuel at about $1.7/gal. The major features of the MARS reactor include sloshing-ion thermal barrier plugs for efficient plasma confinement, a high efficiency blanket, high-field (24-T) choke cells, drift pumping for trapped plasma species, quasi-optical electron-cyclotron resonant heating (ECRH) systems, and a component gridless direct converter.

  19. Mirror Advanced Reactor Study (MARS): executive summary and overview

    International Nuclear Information System (INIS)

    Two self-consistent MARS configurations are discussed - a 1200-MWe commercial electricity-generating plant and a synguels-generating plant that produces hydrogen with an energy equivalent to 26,000 barrels of oil per day. The MARS machine emphasizes the attractive features of the tandem mirror concept, including steady-state operation, a small-diameter high-beta plasma, a linear central cell with simple low-maintenance blankets, low first-wall heat fluxes (2), no driven plasma currents or associated disruptions, natural halo impurity diversion, and direct conversion of end-loss charged-particle power. The MARS electric plant produces 2600 MW of fusion power in a 130-m-long central cell. Advanced tandem-mirror plasma-engineering concepts, a high-efficiency liquid lithium-lead (Li17Pb83) blanket, and efficient direct electrical conversion of end loss power combine to produce a high net plant efficiency of 36%. With a total capital cost of $2.9 billion (constant 1983 dollars), the MARS electric plant produces busbar electricity at approx. 7 cents/kW-hour. The MARS synfuels plant produces 3500 MW of fusion power in a 150-m-long central cell. A helium-gas-cooled silicon carbide pebble-bed blanket provides high-temperature (10000C) heat to a thermochemical water-splitting cycle and the resulting hydrogen is catalytically converted to methanol for distribution. With a total capital cost of $3.6 billion (constant 1983 dollars), the synfuels plant produces methanol fuel at about $1.7/gal. The major features of the MARS reactor include sloshing-ion thermal barrier plugs for efficient plasma confinement, a high efficiency blanket, high-field (24-T) choke cells, drift pumping for trapped plasma species, quasi-optical electron-cyclotron resonant heating (ECRH) systems, and a component gridless direct converter

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

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-13

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

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

    International Nuclear Information System (INIS)

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

  3. 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

  4. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  7. Advanced fuel cycle scenario study in the European context using different burner reactor concepts

    International Nuclear Information System (INIS)

    Different types of fast spectrum dedicated burners have been proposed for the management of radioactive wastes in the frame of various advanced fuel cycle scenarios. Accelerator-driven systems (ADS) and critical low conversion ratio fast reactors have been studied, e.g. within the European context. A potential alternative system is a fusion-fission hybrid (FFH). In the present study, a sodium-cooled fast reactor driven by a D-T fusion neutron source, the subcritical advanced burner reactor (SABR) system is considered. In order to intercompare the different systems, a systematic study is under way. The performances of the two types of systems (SABR, ADS) will be compared from a minor actinide (MA) or transuranic (i.e. Pu+MA) burning potential point of view. The present paper reports preliminary results of the first phase of study, i.e. the comparison of SABR and ADS when used as minor actinides burners. (authors)

  8. Mirror Advanced Reactor Study (MARS) plasma heating systems

    International Nuclear Information System (INIS)

    Plasma heating systems are critical to the operation of a high Q tandem mirror reactor. Two electron cyclotron resonance heating (ECRH) systems, a high energy negative ion-based neutral beam, and an anchor ion cyclotron resonance heating (ICRH) system create the proper axial density, pressure and potential profiles in the endcells. The proper profiles electrostatically plug the central cell plasma, and provide an adequate margin with respect to MHD- and micro-stability for the device. An ICRH system in the central cell is also required to heat the plasma to ignition. Each MARS plasma heating system is at a higher power than that has yet been utilized in any fusion experiments; the use of innovative concepts has lead to compact, efficient, and affordable ECRH and negative beam systems. (orig.)

  9. Sensitivity studies on nuclear data for thorium fuelled Advanced Heavy Water Reactor (AHWR)

    International Nuclear Information System (INIS)

    Sensitivity studies and uncertainty analyses on safety parameters for reactors are an important analysis tool for qualifying the basic nuclear cross-section data. It is also helpful in providing adequate margins at the design stage. In India, the design on Advanced Heavy water Reactor (AHWR) based on thorium is in its advanced stage of development. It is a first-of-a-kind reactor designed with many passive safety features which required to be qualified. In this paper, we discuss several types of sensitivity studies taken up for the integral parameters and reactivity coefficients for the AHWR-reference and the AHWR-LEU variant. The uncertainty studies are required to be taken to level higher where covariances can be established. It is important to analyse the uncertainties in a more rigorous manner

  10. Final report on study of advanced fission power reactor development for the United States. Volume III

    International Nuclear Information System (INIS)

    This three-volume set details a multistage research study on the comparison of various advanced fission reactors and evaluation of alternate strategies for their development in the USA through the year 2020. Volume III presents the basic data and other input information utilized in the process of the study. Detailed reactor and fuel-cycle information is contained on the HTGR, GCFR, MSBR, LWBR, and CANDU-HWR, as obtained by BCL staff from numerous sources. These data have been assembled, critically reviewed, and modified where necessary to assure consistency, both internally for each reactor system, and between reactor types. Similar, comparative information was required for the LWR mass balances, fuel cycle, etc., but this was generally accepted from ERDA sources without the critical evaluation to which the other data were subjected. Other detailed information that relates to the objective of the study, e.g., projections of nuclear power growth and uranium availability, generic safeguards information, description of computerized models, etc., is incorporated for ready reference

  11. A design assessment of tritium removal systems for the mirror advanced reactor study

    International Nuclear Information System (INIS)

    This study investigates the available processes for removing tritium from light water, and selects the most appropriate process for recovering tritium from the various tritiated water streams identified in the Mirror Advanced Reactor Study (MARS). A simplified flowsheet is shown for the process and the main process parameters are identified. Previous experience is utilized to predict direct capital costs and power requirement for the Tritiated Water Removal Unit (TWRU). A number of possibilities are discussed for lowering the cost of the TWRU. An estimate is made of the direct capital cost for the Air Detritiation System that has already been selected as the reference design by MARS personnel. The leakage from the MARS coolant loop is estimated, based on the experience obtained with Ontario Hydro's coolant systems. Design targets are identified for tritium levels in the reactor hall atmosphere and in water and air emissions. Tritium levels are predicted for these and are assessed against the previously identified targets

  12. Study and development of advanced control techniques for nuclear reactors and robots

    Energy Technology Data Exchange (ETDEWEB)

    March-Leuba, C.

    1989-08-01

    This report studies and develops some aspects of the optimal control theory with the objective of evaluating benefits that the nuclear industry could obtain by applying advanced control techniques. First, the basic relationship between optimal control theory and closed-loop control design has been identified. As a result of this work, new algorithms have been developed for feedback implementations. The applicability of these new algorithms to problems such as state estimation, filtering, model update, and model decoupling has been studied. In addition, new alternatives to control design that are not based on optimal control theory have been developed. A broad range of application examples has been presented for several physical systems, including pressurized water nuclear reactors, boiling water nuclear reactors, steam generators, and robotics. 22 refs., 26 figs.

  13. Study and development of advanced control techniques for nuclear reactors and robots

    International Nuclear Information System (INIS)

    This report studies and develops some aspects of the optimal control theory with the objective of evaluating benefits that the nuclear industry could obtain by applying advanced control techniques. First, the basic relationship between optimal control theory and closed-loop control design has been identified. As a result of this work, new algorithms have been developed for feedback implementations. The applicability of these new algorithms to problems such as state estimation, filtering, model update, and model decoupling has been studied. In addition, new alternatives to control design that are not based on optimal control theory have been developed. A broad range of application examples has been presented for several physical systems, including pressurized water nuclear reactors, boiling water nuclear reactors, steam generators, and robotics. 22 refs., 26 figs

  14. A CFD Modeling Study for the Design of an Advanced HANARO Reactor Core Structure

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong-Hark; Chae, Hee-Teak; Park, Cheol; Kim, Heo-Nil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    AHR(Advanced HANARO Reactor) based on HANARO has been under a conceptually designed with new ideas to implement new findings, which have been revealed from twelve years operation of HANARO. For example, a perforated structure to reduce the FIV(Flow Induced Vibration) of a fuel assembly has been considered to install. And a change of dual outlets to a single outlet has also been investigated to promote the accessibility and to work easily in the reactor pool. Those investigations have been conducted by the CFD (Computational Fluid Dynamics) method, which can provide us with an good understanding of three dimensional flow fields influenced by design changes without an experiment. In this study a CFD modeling study for an AHR core structure design is described.

  15. A CFD Modeling Study for the Design of an Advanced HANARO Reactor Core Structure

    International Nuclear Information System (INIS)

    AHR(Advanced HANARO Reactor) based on HANARO has been under a conceptually designed with new ideas to implement new findings, which have been revealed from twelve years operation of HANARO. For example, a perforated structure to reduce the FIV(Flow Induced Vibration) of a fuel assembly has been considered to install. And a change of dual outlets to a single outlet has also been investigated to promote the accessibility and to work easily in the reactor pool. Those investigations have been conducted by the CFD (Computational Fluid Dynamics) method, which can provide us with an good understanding of three dimensional flow fields influenced by design changes without an experiment. In this study a CFD modeling study for an AHR core structure design is described

  16. Analytical modelling and study of the stability characteristics of the Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    An analytical model has been developed to study the thermohydraulic and neutronic-coupled density-wave instability in the Indian Advanced Heavy Water Reactor (AHWR) which is a natural circulation pressure tube type boiling water reactor. The model considers a point kinetics model for the neutron dynamics and a lumped parameter model for the fuel thermal dynamics along with the conservation equations of mass, momentum and energy and equation of state for the coolant. In addition, to study the effect of neutron interactions between different parts of the core, the model considers a coupled multipoint kinetics equation in place of simple point kinetics equation. Linear stability theory was applied to reveal the instability of in-phase and out-of-phase modes in the boiling channels of the AHWR. The results indicate that the design configuration considered may experience both Ledinegg and Type I and Type II density-wave instabilities depending on the operating condition. Some methods of suppressing these instabilities were found out. In addition, it was found that the stability behavior of the reactor is greatly influenced by the void reactivity coefficient, fuel time constant, radial power distribution and channel inlet orificing. The delayed neutrons were found to have strong influence on the Type I and Type II instabilities. Decay ratio maps were predicted considering various operating parameters of the reactor, which are useful for its design. (author)

  17. Evaluation of trial design studies for an advanced marine reactor, (7)

    International Nuclear Information System (INIS)

    We have performed the numerical evaluation of shielding design for three advanced marine reactors (semi-integral PWR, integral PWR and self-pressurized PWR) under operational condition and hypothetical accident. Common calculational procedure and shielding material ingredient over the three reactors have been adopted for fair evaluation. (author)

  18. 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)

  19. 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

  20. Thermal hydraulic studies for passive heat transport systems relevant to advanced reactors

    International Nuclear Information System (INIS)

    Nuclear is the only non-green house gas generating power source that can replace fossil fuels and can be commercially deployed in large scale. However, the enormous developmental efforts and safety upgrades during the past six decades have somewhat eroded the economic competitiveness of water-cooled reactors which form the mainstay of the current nuclear power programme. Further, the introduction of the supercritical Rankine cycle and the gas turbine based advanced fuel cycles have enhanced the efficiency of fossil fired power plants (FPP) thereby reducing its greenhouse gas emissions. The ongoing development of ultra-supercritical and advanced ultra-supercritical turbines aims to further reduce the greenhouse gas emissions and economic competitiveness of FPPs. In the backdrop of these developments, the nuclear industry also initiated development of advanced nuclear power plants (NPP) with improved efficiency, sustainability and enhanced safety as the main goals. A review of the advanced reactor concepts being investigated currently reveals that excepting the SCWR, all other concepts use coolants other than water. The coolants used are lead, lead bismuth eutectic, liquid sodium, molten salts, helium and supercritical water. Besides, some of these are employing passive systems to transport heat from the core under normal operating conditions. In view of this, a study is in progress at BARC to examine the performance of simple passive systems using SC CO2, SCW, LBE and molten salts as the coolant. This paper deals with some of the recent results of these studies. The study focuses on the steady state, transient and stability behaviour of the passive systems with these coolants. (author)

  1. Interim results of the study of control room crew staffing for advanced passive reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, B.P.; Sebok, A.; Haugset, K. [OECD Halden Reactor Project (Norway)

    1996-03-01

    Differences in the ways in which vendors expect the operations staff to interact with advanced passive plants by vendors have led to a need for reconsideration of the minimum shift staffing requirements of licensed Reactor Operators and Senior Reactor Operators contained in current federal regulations (i.e., 10 CFR 50.54(m)). A research project is being carried out to evaluate the impact(s) of advanced passive plant design and staffing of control room crews on operator and team performance. The purpose of the project is to contribute to the understanding of potential safety issues and provide data to support the development of design review guidance. Two factors are being evaluated across a range of plant operating conditions: control room crew staffing; and characteristics of the operating facility itself, whether it employs conventional or advanced, passive features. This paper presents the results of the first phase of the study conducted at the Loviisa nuclear power station earlier this year. Loviisa served as the conventional plant in this study. Data collection from four crews were collected from a series of design basis scenarios, each crew serving in either a normal or minimum staffing configuration. Results of data analyses show that crews participating in the minimum shift staffing configuration experienced significantly higher workload, had lower situation awareness, demonstrated significantly less effective team performance, and performed more poorly as a crew than the crews participating in the normal shift staffing configuration. The baseline data on crew configurations from the conventional plant setting will be compared with similar data to be collected from the advanced plant setting, and a report prepared providing the results of the entire study.

  2. Core and Refueling Design Studies for the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Ilas, Dan [ORNL; Varma, Venugopal Koikal [ORNL; Cisneros, Anselmo T [ORNL; Kelly, Ryan P [ORNL; Gehin, Jess C [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure

  3. Numerical Study on Crossflow Printed Circuit Heat Exchanger for Advanced Small Modular Reactors

    International Nuclear Information System (INIS)

    Various fluids such as water, gases (helium), molten salts (FLiNaK, FLiBe) and liquid metal (sodium) are used as a coolant of advanced small modular reactors (SMRs). The printed circuit heat exchanger (PCHE) has been adopted as the intermediate and/or secondary heat exchanger of SMR systems because this heat exchanger is compact and effective. The size and cost of PCHE can be changed by the coolant type of each SMR. In this study, the crossflow PCHE analysis code for advanced small modular reactor has been developed for the thermal design and cost estimation of the heat exchanger. The analytical solution of single pass, both unmixed fluids crossflow heat exchanger model was employed to calculate a two dimensional temperature profile of a crossflow PCHE. The analytical solution of crossflow heat exchanger was simply implemented by using built in function of the MATLAB program. The effect of fluid property uncertainty on the calculation results was evaluated. In addition, the effect of heat transfer correlations on the calculated temperature profile was analyzed by taking into account possible combinations of primary and secondary coolants in the SMR systems. Size and cost of heat exchanger were evaluated for the given temperature requirement of each SMR

  4. Study of plutonium disposition using existing GE advanced Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

  5. Study of plutonium disposition using existing GE advanced Boiling Water Reactors

    International Nuclear Information System (INIS)

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ''Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs

  6. Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-04-30

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.

  7. 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...

  8. 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...

  9. Study of an advanced D-T tokamak fusion reactor with compact fusion advanced rankine (CFAR) cycle

    International Nuclear Information System (INIS)

    Recent progress of the CFAR (Compact Fusion Advanced Rankine) cycle concept for a D-T tokamak reactor is presented with emphasis on how an enthalpy extraction can be achieved by a nonequilibrium disk-type MHD generator. For the gas stagnation temperatures of 3,000 K, enthalpy extraction in excess of 50% is found to be achievable, leading to a 40% overall plant efficiency with application of recuperative heat cycle and recently advanced thermoelectric converters. About 6 ton/sec mercury flow is required to remove fusion energy while achieving the 3,000 K gas stagnation temperature prior to the MHD generator. Studies of plasma parameters in the steady-state operation regime subject to plasma physics constraints, the minimum power in the start up phase required for ignition, effects of MHD magnet to the plasma confining magnetic fields, neutron and microwave superheat, and mercury corrosion test of ceramic rods for 2,000 hours are also described. 14 refs., 6 figs., 1 tab

  10. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-09-15

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  11. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    International Nuclear Information System (INIS)

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval

  12. Fast Reactor Fuel Cycle Cost Estimates for Advanced Fuel Cycle Studies

    International Nuclear Information System (INIS)

    Presentation Outline: • Why Do I Need a Cost Basis?; • History of the Advanced Fuel Cycle Cost Basis; • Description of the Cost Basis; • Current Work; • Fast Reactor Fuel Cycle Applications; • Sample Fuel Cycle Cost Estimate Analysis; • Future Work

  13. 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

  14. Studies on advanced water-cooled reactors beyond generation Ⅲ for power generation

    Institute of Scientific and Technical Information of China (English)

    CHENG Xu

    2007-01-01

    China's ambitious nuclear power program motivates the country's nuclear community to develop advanced reactor concepts beyond generation Ⅲ to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics,sustainability and technology availability. It is a logical extension of the generation Ⅲ PWR technology in China.The status of international R&D work is reviewed. A new supercritieal water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydranlics method is carded out. It shows good feasibility for the new design proposal.

  15. 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)

  16. 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

  17. Study on recriticality of fuel debris during hypothetical severe accidents in the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    A study has been performed to measure the potential of recriticality during hypothetical severe accident in Advanced Neutron Source (ANS). For the lumped debris configuration in the Reactor Coolant System (RCS), as found in the previous study, recriticality potential may be very low. However, if fuel debris is dispersed and mixed with heavy water in RCS, recriticality potential has been predicted to be substantial depending on thermal-hydraulic conditions surrounding fuel debris mixture. The recriticality potential in RCS is substantially reduced for the three element core design with 50% enrichment. Also, as observed in the previous study, strong dependencies of keff on key thermal hydraulic parameters are shown. Light water contamination is shown to provide a positive reactivity, and void formation due to boiling of mixed water provides enough negative reactivity and to bring the system down to subcritical. For criticality potential in the subpile room, the lumped debris configuration does not pose a concern. Dispersed configuration in light water pool of the subpile room is also unlikely to result in criticality. However, if the debris is dispersed in the pool that is mixed with heavy water, the results indicate that a substantial potential exists for the debris to reach the criticality. However, if prompt recriticality disperses the debris completely in the subpile room pool, subsequent recriticality may be prevented since neutron leakage effects become large enough

  18. The study of steam explosions in nuclear systems. Advanced Reactor Severe Accident Program

    International Nuclear Information System (INIS)

    This report presents an overview of the steam explosion issue in nuclear reactor safety and our approach to assessing it. Key physics, models, and computational tools are described, and illustrative results are presented for ex-vessel steam explosions in an open pool geometry. An extensive set of appendices facilitate access to previously reported work that is an integral part of this effort. These appendices include key developments in our approach, key advances in our understanding from physical and numerical experiments, and details of the most advanced computational results presented in this report. Of major significance are the following features: A consistent two-dimensional treatment for both premixing and propagation which in practical settings are ostensibly at least two-dimensional phenomena; experimental demonstration of voiding and microinteractions which represent key behaviors in premixing and propagation respectively; demonstration of the explosion venting phenomena in open pool geometries which, therefore, can be counted on as a very important mitigative feature; and introduction of the idea of penetration cutoff as a key mechanism prohibiting large-scale premixing in usual ex-vessel situations involving high pour velocities and subcooled pools. This report is intended as an overview and is to be followed by code manuals for PM-ALPHA and ESPROSE.m, respective verification reports, and application documents for reactor-specific applications. The applications will employ the Risk Oriented Accident Analysis Methodology (ROAAM) to address the safety importance of potential steam explosions phenomena in evaluated severe accidents for passive Advanced Light Water Reactors (ALWRs)

  19. 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

  20. 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)

  1. A study of the advancement of a reactor core design environment

    International Nuclear Information System (INIS)

    Full text: During the years from 2002 to 2004 a joint project has been performed by IFE, Halden and Yonden Engineering Corporation, Japan, to develop an advanced reactor core design environment based on a communication method for controlling a reactor core code system efficiently from PCs in a distributed network. The advanced reactor core design environment is realized by using Microsoft Visual Basic and communication software based on the IFE product SoftwareBus. The project has been carried out based on the fact that a computer-aided design system has been under development at Yonden Engineering Corporation in order to perform efficiently fuel replacement calculation by Yonden's reactor design code system. In this system, the structure is such that the physics calculation code system runs on UNIX workstations (in parallel) performing the calculations, while the Man-Machine Interface for controlling the calculation programs run on PCs in a distributed network. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general communication tool (IFE's SoftwareBus) has been used for realizing communication of the n-pair n-node between the reactor core design code system and the PC applications. Further, a method of improvement in the speed of the optimal pattern calculation has been implemented by assigning each examination pattern to two or more computers distributed in the network and assigning the next pattern calculation to the computer, where the calculation has ended or has the lowest workload. The high-speed technology of the pattern survey by network distributed processing is based on SoftwareBus. The reactor core design code system is developed in FORTRAN running on a UNIX workstation (Solaris). The PC applications have been developed by using Microsoft Visual Basic on Windows 2000 platform. The first step of the verification and validation process was carried out in March

  2. Experimental studies on in-bundle ECCS injection for Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    The Advanced Heavy Water Reactor (AHWR) being designed at BARC is an innovative reactor with Thorium utilization as its major objective. It has many advanced passive safety features. One such feature is passive injection of emergency coolant after postulated Loss of Coolant Accident (LOCA). A novel feature of this injection scheme is that the injection does not take place in the header/plenum as in other reactors, but directly in to the bundle. For this purpose, the fuel cluster incorporates a central water rod which communicates with the ECCS header. The water rod extends along full length of the fuel cluster. In event of LOCA in the Main Heat Transport (MHT) system, ECC water flows from the accumulator to the water rod through ECCS header. The water flows into the bundle through holes in the water rod. The AHWR fuel cluster has fuel pins arranged in three concentric rings (of 12, 18 and 24 pins) around the central rod. While it is ensured that water does reach the fuel cluster, whether it reaches the outer ring of pins is needs investigation as the pins are closely spaced (1-3 mm gap between adjacent rods). The objective of the present experiments is to determine under what conditions (ECC flow and decay heat), the ECC water is able to rewet and cool all the fuel pins. The experiments have been done in a short, instrumented fuel bundle simulating the geometry of the AHWR fuel cluster

  3. 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

  4. 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.

  5. Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

    2012-08-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

  6. 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

  7. 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

  8. Advanced reactors in France

    International Nuclear Information System (INIS)

    In 1980, in an attempt to reduce costs, and due to the coming to an end of the licence agreements between FRAMATOME and WESTINGHOUSE, the study of a new standardized plants series was undertaken. It aimed at deriving profit from the lessons drawn from the experience feedback and at taking into account the evolution of the regulations background, and an acute awareness of the necessity to improve the ''man-machine'' interface. However, one should already pay attention to what will replace the N4 standardized plant series and the obsolete nuclear power plants, at the beginning of the next century

  9. Passive Safety Systems in Advanced Water Cooled Reactors (AWCRS). Case Studies. A Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    This report presents the results from the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) collaborative project (CP) on Advanced Water Cooled Reactor Case Studies in Support of Passive Safety Systems (AWCR), undertaken under the INPRO Programme Area C. INPRO was launched in 2000 - on the basis of a resolution of the IAEA General Conference (GC(44)/RES/21) - to ensure that nuclear energy is available in the 21st century in a sustainable manner, and it seeks to bring together all interested Member States to consider actions to achieve innovation. An important objective of nuclear energy system assessments is to identify 'gaps' in the various technologies and corresponding research and development (R and D) needs. This programme area fosters collaboration among INPRO Member States on selected innovative nuclear technologies to bridge technology gaps. Public concern about nuclear reactor safety has increased after the Fukushima Daiichi nuclear power plant accident caused by the loss of power to pump water for removing residual heat in the core. As a consequence, there has been an increasing interest in designing safety systems for new and advanced reactors that are passive in nature. Compared to active systems, passive safety features do not require operator intervention, active controls, or an external energy source. Passive systems rely only on physical phenomena such as natural circulation, thermal convection, gravity and self-pressurization. Passive safety features, therefore, are increasingly recognized as an essential component of the next-generation advanced reactors. A high level of safety and improved competitiveness are common goals for designing advanced nuclear power plants. Many of these systems incorporate several passive design concepts aimed at improving safety and reliability. The advantages of passive safety systems include simplicity, and avoidance of human intervention, external power or signals. For these reasons, most

  10. Study for requirement of advanced long life small modular fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tak, Taewoo, E-mail: ttwispy@unist.ac.kr; Choe, Jiwon, E-mail: chi91023@unist.ac.kr; Jeong, Yongjin, E-mail: yjjeong09@unist.ac.kr; Lee, Deokjung, E-mail: deokjung@unist.ac.kr [Ulsan National Institute of Science and Technology, 50, UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan, 689-798 (Korea, Republic of); Kim, T. K., E-mail: tkkim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60564 (United States)

    2016-01-22

    To develop an advanced long-life SMR core concept, the feasibility of the long-life breed-and-burn core concept has been assessed and the preliminary selection on the reactor design requirement such as fuel form, coolant material has been performed. With the simplified cigar-type geometry of 8m-tall CANDLE reactor concept, it has demonstrated the strengths of breed-and-burn strategy. There is a saturation region in the graph for the multiplication factors, which means that a steady breeding is being proceeded along the axial direction. The propagation behavior of the CANDLE core can be also confirmed through the evolution of the axial power profile. Coolant material is expected to have low melting point, density, viscosity and absorption cross section and a high boiling point, specific heat, and thermal conductivity. In this respect, sodium is preferable material for a coolant of this nuclear power plant system. The metallic fuel has harder spectrum compared to the oxide and carbide fuel, which is favorable to increase the breeding and extend the cycle length.

  11. Study for requirement of advanced long life small modular fast reactor

    Science.gov (United States)

    Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung; Kim, T. K.

    2016-01-01

    To develop an advanced long-life SMR core concept, the feasibility of the long-life breed-and-burn core concept has been assessed and the preliminary selection on the reactor design requirement such as fuel form, coolant material has been performed. With the simplified cigar-type geometry of 8m-tall CANDLE reactor concept, it has demonstrated the strengths of breed-and-burn strategy. There is a saturation region in the graph for the multiplication factors, which means that a steady breeding is being proceeded along the axial direction. The propagation behavior of the CANDLE core can be also confirmed through the evolution of the axial power profile. Coolant material is expected to have low melting point, density, viscosity and absorption cross section and a high boiling point, specific heat, and thermal conductivity. In this respect, sodium is preferable material for a coolant of this nuclear power plant system. The metallic fuel has harder spectrum compared to the oxide and carbide fuel, which is favorable to increase the breeding and extend the cycle length.

  12. Study for requirement of advanced long life small modular fast reactor

    International Nuclear Information System (INIS)

    To develop an advanced long-life SMR core concept, the feasibility of the long-life breed-and-burn core concept has been assessed and the preliminary selection on the reactor design requirement such as fuel form, coolant material has been performed. With the simplified cigar-type geometry of 8m-tall CANDLE reactor concept, it has demonstrated the strengths of breed-and-burn strategy. There is a saturation region in the graph for the multiplication factors, which means that a steady breeding is being proceeded along the axial direction. The propagation behavior of the CANDLE core can be also confirmed through the evolution of the axial power profile. Coolant material is expected to have low melting point, density, viscosity and absorption cross section and a high boiling point, specific heat, and thermal conductivity. In this respect, sodium is preferable material for a coolant of this nuclear power plant system. The metallic fuel has harder spectrum compared to the oxide and carbide fuel, which is favorable to increase the breeding and extend the cycle length

  13. 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.

  14. 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.

  15. 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

  16. Reduction of Worldwide Plutonium Inventories Using Conventional Reactors and Advanced Fuels: A Systems Study

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

    The potential for reducing plutonium inventories in the civilian nuclear fuel cycle through recycle in LWRs of a variety of mixed oxide forms is examined by means of a cost based plutonium flow systems model. This model emphasizes: (1) the minimization of separated plutonium; (2) the long term reduction of spent fuel plutonium; (3) the optimum utilization of uranium resources; and (4) the reduction of (relative) proliferation risks. This parametric systems study utilizes a globally aggregated, long term (approx. 100 years) nuclear energy model that interprets scenario consequences in terms of material inventories, energy costs, and relative proliferation risks associated with the civilian fuel cycle. The impact of introducing nonfertile fuels (NFF,e.g., plutonium oxide in an oxide matrix that contains no uranium) into conventional (LWR) reactors to reduce net plutonium generation, to increase plutonium burnup, and to reduce exo- reactor plutonium inventories also is examined.

  17. Study of advanced nuclear fuel cycles in Candu type power reactors

    International Nuclear Information System (INIS)

    The fuel burn up can be increased to a large extent, up to 14, 0000 MWD/te, by using the slightly enriched uranium or Pu mixed fuel in CANDU type power reactors. In the present study, the previous work was extended to compare the isotopic inventories and corresponding activities of important nuclides for different fuel cycles of a CANDU 600 type power reactor. The detail can be found in our studies. The calculations were performed using the computer code WIMSD4. The isotopic inventories and corresponding activities were calculated versus the fuel burn-up for the natural UO/sub 2/ fuel, 1.2 % enriched UO/sub 2/ fuel and 0.45 % PuO/sub 2/-UO/sub 2/ fuel. It was found that 1.2 % enriched uranium fuel has the lowest activity as compared to other two fuel cycles. It means that improvement in the fuel cycle technology of CANDU type power reactors can lead to high burn up which results in the reduction of actinide content in the spent fuel, and hence has a good environmental impact. (orig./A.B.)

  18. Intermediate Heat Exchanger (IHX) Trade Study in Support of Advanced Fast Reactor-100 Development

    International Nuclear Information System (INIS)

    Reactor fabrication costs are closely linked to the commodities utilized in their construction. One large component of many pool-type sodium fast reactor designs that can significantly impact the overall size of the reactor vessel is the Intermediate Heat Exchanger (IHX). On this basis, a trade study was performed to evaluate four different types of novel IHX designs from the viewpoint of minimizing the impact on the size of the reactor. The four designs that were evaluated as part of this study were: i) conventional shell-with-straight tube design, ii) shell with twisted tube design, iii) Printed Circuit Heat Exchanger (PCHE), and iv) a spiral tube design. The approach was to develop simple thermal-hydraulics models in order to provide a technical basis for assessing the overall size of the units. Simple mechanical stress models were also developed so that component thicknesses could be calculated. This allowed material volumes and commodities for the various designs to be estimated. The results indicate that adopting a twisted vs. straight tube design can reduce commodities by ~ 30 %, and overall height of the units by roughly 30-40%. Aside from the choice of tube design, geometry also plays an important role in reducing commodity and space requirements; i.e., commodities for a kidney-shaped design were found to be ~10% less relative to the circular case, and the reduction is even more (~20%) for an annular design. If a kidney-shaped design is adopted, then the radial thickness of the vessel occluded by the IHXs can be reduced by 20%. Conversely, the reduction for annular designs approaches 75%. Height requirements are also reduced, and this is principally due to the smaller radius of the elliptical heads that are required in the radially thinned designs. Additional costs that may be incurred by fabricating a more complex design have not been factored into these analyses. (author)

  19. 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)

  20. 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

  1. Nonlinear stability analysis of a reduced order model of nuclear reactors: A parametric study relevant to the advanced heavy water reactor

    International Nuclear Information System (INIS)

    Research highlights: → We model power oscillations in boiling water reactors using a lumped parameter model. → The nature and amplitudes of oscillations is obtained using a nonlinear analysis. → The method of multiple scales has been used for the analytical treatment. → Fuel temperature coefficient of reactivity determines the nature of oscillations. → The presented systematic method of analysis useful for reduced order reactor models. - Abstract: In this paper, we perform a parametric study of the nonlinear dynamics of a reduced order model for boiling water reactors (BWR) near the Hopf bifurcation point using the method of multiple scales (MMS). Analysis has been performed for general values of the parameters, but the results are demonstrated for parameter values of the model corresponding to the advanced heavy water reactor (AHWR). The neutronics of the AHWR is modeled using point reactor kinetic equations while a one-node lumped parameter model is assumed both for the fuel and the coolant for modeling the thermal-hydraulics. Nonlinearities in the heat transfer process are ignored and attention is focused on the nonlinearity introduced by the reactivity feedback. It is found that the steady-state operation of the AHWR mathematical model looses stability via. a Hopf bifurcation resulting in power oscillations as some typical bifurcation parameter like the void coefficient of reactivity is varied. The bifurcation is found to be subcritical for the parameter values corresponding to the AHWR. However, with a decrease in the fuel temperature coefficient of reactivity the bifurcation turns to supercritical implying global stability of the steady state operation in the linear stability regime. Moreover slight intrusion into the instability regime results in small-amplitude limit cycles leaving the possibility of retracting back to stable operation.

  2. Study of Advanced Railgun Hydrogen Pellet Injectors for Fusion Reactor Refueling.

    Science.gov (United States)

    King, Tony Levone

    An advanced railgun system has been developed to assess its feasibility as a hypervelocity hydrogen pellet injector for magnetically confined plasmas. It consists of a pellet generator/gas gun assembly for freezing hydrogen pellets and injecting them into the railgun at velocities as high as 1.5 km/s. A plasma armature is formed by ionizing the low-Z propellant gas behind the pellet and firing the railgun. This fuseless operation prevents high-Z impurities from entering the reactor during pellet injection. The railgun system has several features that distinguish it from its predecessors, including: (1) a more compact, versatile pellet generator, (2) a new gas gun configuration that produces significantly higher pellet speeds, (3) a perforated coupling piece between the gas gun and railgun to prevent spurious arcing, and (4) ablation-resistant sidewalls, perforated sidewalls and transaugmentation to reduce inertial and viscous drag, the primary obstacles to achieving hypervelocity. A unique system of sophisticated controls and diagnostics has been assembled to operate the railgun system and assess its performance, including fully automated pellet freezing and gas gun operation, an automatic timing circuit that is immune to mistriggering caused by pellet fragmentation or electromagnetic interference, a streak camera, photostations, light gates, current trans formers, B-dot probes, laser interferometry and optical spectroscopy. Free-arc and hydrogen pellet experiments were conducted to evaluate various railgun designs. Transaugmented and simple railguns 1.2 and 2 m long were tested. The performances of railguns using Mullite, solid Lexan and perforated Lexan sidewalls were compared. The railgun theory of operation and anticipated losses are also examined. The theoretical predictions are found to be in good agreement with the experimental results. The advanced railgun system has set several world records for bare hydrogen pellet velocity, including a 3.3 km/s shot on

  3. The OPERA loop in the OSIRIS reactor core. Pressurized-water irradiation device to study Advanced Reactor Fuels

    International Nuclear Information System (INIS)

    This loop is designed to allow fuel qualification test, i.d. to allow irradiation of representative parts fuel assemblies operating in thermohydraulic and chemical conditions representative of these of present pressurized water reactors or in development. This paper presents the aims of the installation, the general design and the main specifications with a brief detailed description

  4. The EU power plant conceptual study - neutronic design analyses for near term and advanced reactor models

    International Nuclear Information System (INIS)

    A power plant conceptual study (PPCS) has been conducted in the framework of the European fusion programme with the main objective to demonstrate the safety and environmental advantages and the economic viability of fusion power. Power plant models with limited (''near term concepts'') and advanced plasma physics and technological extrapolations (''advanced concepts'') were considered. Two near term plant models were selected, one employing a water cooled lithium-lead (WCLL), and the other one a helium cooled pebble bed (HCPB) blanket. Two variants were also considered for the advanced power plant models, one adopting a liquid metal blanket with a self-cooled lithium-lead breeder zone and a helium cooled steel structure (''dual coolant lithium lead'', DCLL), and the other one a self-cooled lithium-lead (SCLL) blanket with SiCf/SiC composite as structural material. This report provides a detailed documentation of the neutronics design analyses performed as part of the PPCS study for both the near term and advanced power plant models. Main issues are the assessment of the tritium breeding capability, the evaluation of the nuclear power generation and its spatial distribution, and the assessment and optimisation of the shielding performance. The analyses were based on three-dimensional Monte Carlo calculations with the MCNP code using suitable torus sector models developed for the different PPCS plant variants. (orig.)

  5. Fundamental study on thermo-hydraulic phenomena concerning passive safety of advanced marine reactor

    International Nuclear Information System (INIS)

    The objective of this study is to investigate the thermo-hydraulic behavior of a fluid region confined in a rectangular parallelepiped cavity equipped with a heater and a cooler. The motivation of this study is to clarify a thermal buffer effect for an innovative marine nuclear reactor to realize passive safety. In the present study, experiments were carried out with conditions of laminar convection. Temperature and flow behavior was visualized by the liquid-crystal suspension method, by which the temperature distribution in liquid can be observed as a colored map. Thermal plumes from the heater and the cooler, global natural circulation in the cavity and thermal stratification were observed as elements of the complicated phenomena. Using a code which solves the Navier-Stokes and energy equations, numerical simulations under steady and unsteady condition were carried out to predict the experimental results for two-dimensional, laminar situations, and a good agreement was obtained. (author)

  6. 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

  7. 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.)

  8. 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)

  9. 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

  10. Experimental and analytical studies of passive shutdown heat removal from advanced LMRs [liquid metal reactors

    International Nuclear Information System (INIS)

    A facility designed and constructed to demonstrate the viability of natural convection passive heat removal systems as a key feature of innovative LMR Shutdown Heat Removal (SHR) systems is in operation at Argonne National Laboratory (ANL). This Natural Convection Shutdown Heat Removal Test Facility (NSTF) has investigated the heat transfer performance of the GE/PRISM passive design. This initial series of experiments simulates the air-side geometry of the PRISM Radiant Reactor Vessel Auxiliary Cooling System (RVACS). The NSTF operates in either a uniform heat flux mode and a uniform temperature mode at the air/guard vessel interface. Analysis of the RVACS performance data indicates excellent agreement with pretest analytical predictions. Correlation analysis presents the heat transfer data in a form suitable for use in LMR design and verification of analytical studies

  11. Advances of study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    R and D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R and D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R and D plan and describe some advances on experimental and analytical studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility and the analytical one aims to develop a predictable technology for geometry effects such as gap between rods, grid spacer configuration etc. using advanced 3-D two-phase flow simulation methods. Steady-state and transient critical power experiments are conducted with the test facility (Gap width between rods: 1.0 mm) and the experimental data reveal the feasibility of RMWR. (authors)

  12. 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

  13. 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

  14. Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

    International Nuclear Information System (INIS)

    Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations

  15. Reduction of worldwide plutonium inventories using conventional reactors and advanced fuels. A systems study

    International Nuclear Information System (INIS)

    The potential for reducing plutonium inventories in the civilian nuclear fuel cycle through recycle in LWRs is examined by means of a cost-based plutonium-flow systems model that includes an approximate measure of proliferation risk. The impact of plutonium recycle forms is examined, including the introduction of nonfertile fuels into conventional (LWR) reactors to reduce net plutonium generation, to increase plutonium burnup, and to reduce exo-reactor plutonium inventories. (author)

  16. Study on advanced reinforced concrete containment for the next generation boiling water reactors

    International Nuclear Information System (INIS)

    The paper presents the study results about advanced reinforced concrete containment (ARCCV) for the next generation BWRs. The concept is general and applicable to all BWRs which have internal recirculation pumps (RIP). The ARCCV is based on the current reinforced concrete containment vessel (RCCV) which is used for the current ABWR. There are three major improvements from the current RCCV. These improvements enables the reduction of the R/B volume and improvement of safety simultaneously. The first improvement of the ARCCV is the separation between the upper DW and the lower DW. This contributes to the reduction of the nitrogen and hydrogen transferred to the WW air space at an accident. The pressure increase by the compression in the WW air space can be minimized and hence the reduction of the peak pressure of the containment dramatically regardless of the power increase to) 1,700 MWe. The second improvement of the ARCCV is the adoption of the advanced passive cooling system (APCS). It can minimize the necessity of the containment venting at severe accidents. The third improvement of the ARCCV is the raised S/P. Most of the ECCS pumps can be installed under the raised S/P. This can reduce the area of the lowest floor of the R/B and hence the reduction of the R/B volume. The above three improvements enables the cost saving of the R/B and safety improvements for both DBA LOCA and severe accidents dramatically. (author)

  17. 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

  18. Severe accident risk minimization studies for the Advanced Neutron Source (ANS) reactor plant at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    This paper discusses salient aspects of severe accident related phenomenological considerations, scoping studies, and mitigative design features being studied for incorporation into a high-power research reactor plant. Key results of scoping studies on steam explosions, recriticality, core-concrete interactions, and containment transport are highlighted. Evolving design features of the containment are described. Containment response calculations for a site-suitability basis transient are presented that demonstrate acceptable source term values and superior containment performance. Oak Ridge National Laboratory's (ORNL) Advanced Neutron Source (ANS) will be a new user facility for all kinds of neutron research, centered around a research reactor of unprecedented neutron beam flux. A defense-in-depth philosophy has been adopted. In response to this commitment, ANS Project management initiated severe accident analysis and related technology development early-on in the design phase itself. This was done to aid in designing a sufficiently robust containment for retention and controlled release of radionuclides in the event of such an accident. It also provides a means for satisfying on- and off-site regulatory requirements, accident-related dose exposures, and containment response and source-term best-estimate analyses for level-2 and -3 Probabilistic Risk Analysis (PRAs) that will be produced. Moreover, it will provide the best possible understanding of the ANS under severe accident conditions and consequently provide insights for the development of strategies and design philosophies for accident mitigation, management, and emergency preparedness efforts

  19. Reduction of worldwide plutonium inventories using conventional reactors and advanced fuels: A systems study

    International Nuclear Information System (INIS)

    The potential for reducing plutonium inventories in the civilian nuclear fuel cycle through recycle in LWRs of a variety of mixed-oxide forms is examined by means of a cost-based plutonium-flow systems model that includes an approximate measure of proliferation risk. The impact of plutonium recycle in a number of forms is examined, including the introduction of nonfertile fuels into conventional (LWR) reactors to reduce net plutonium generation, to increase plutonium burnup, and to reduce exo-reactor plutonium inventories

  20. 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

  1. Mirror Advanced Reactor Study (MARS). Final report. Volume 2. Commercial fusion synfuels plant

    International Nuclear Information System (INIS)

    Volume 2 contains the following chapters: (1) synfuels; (2) physics base and parameters for TMR; (3) high-temperature two-temperature-zone blanket system for synfuel application; (4) thermochemical hydrogen processes; (5) interfacing the sulfur-iodine cycle; (6) interfacing the reactor with the thermochemical process; (7) tritium control in the blanket system; (8) the sulfur trioxide fluidized-bed composer; (9) preliminary cost estimates; and (10) fuels beyond hydrogen

  2. 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

  3. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

    -plates will be conducted in the Advanced Test Reactor(ATR). 49 compacts with a uranium density of 8 gU/cc consist of 7 different atomized uranium-molybdenum alloy powders. The tensile strength increased and the elongation decreased with increasing the volume fraction of U-10Mo powders in dispersion fuel. The tensile strength was lower and elongation was larger in dispersion fuel using atomized U-10Mo powders than that using comminuted fuel powders. The green strength of the comminuted powder compacts was about twice as large as that of the atomized powder compacts. It is suggested that the compacting condition required to fabricate the atomized powder compacts is over the 350MPa. The comminuted irregular shaped particles and smaller particle size of fuel powders showed improved homogeneity of powder mixture. The homogeneity of powder mixtures increased to a minimum at approximately 0.10 wt% moisture and then decreased with moisture content.

  4. Results of the IAEA CRP on studies of advanced reactor technology options for effective incineration of radioactive waste

    International Nuclear Information System (INIS)

    The IAEA has initiated a Coordinated Research Project (CRP) on 'Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste'. The overall objective of the CRP, performed within the framework of IAEA's Nuclear Power Technology Development Section's Technical Working Group on Fast Reactors (TWG-FR), is to increase the capability of Member States in developing and applying advanced technologies in the area of long-lived radioactive waste utilization and transmutation. More specifically, the final goal of the CRP is to deepen the understanding of the dynamics of transmutation systems, especially systems with high minor actinide content. Currently, 20 institutions from 15 member states and one international organization are participating in this CRP. The current author list comprises the participants of the last CRP Vienna meeting. The CRP concentrates on the assessment of the transient behaviour of various transmutation systems. For a sound assessment of the transient and accident behaviour, neutron kinetics and dynamics methods and codes have to be qualified, especially as the margins for the safety relevant neutronics parameters are generally becoming small in a transmutation system. Hence, the availability of adequate and qualified methods for the analysis of the various systems is an important point of the exercise. A benchmarking effort between the codes and nuclear data used for the analyses has been performed, which will help specifying the range of validity of methods, and also formulate requirements for future theoretical and experimental research. Should transient experiments become available during the course of the CRP, experimental benchmarking work will also be pursued

  5. 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.)

  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. 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

  8. 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.)

  9. 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

  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. Study of Cost Effective Large Advanced Pressurized Water Reactors that Employ Passive Safety Features

    International Nuclear Information System (INIS)

    A report of DOE sponsored portions of AP1000 Design Certification effort. On December 16, 1999, The United States Nuclear Regulatory Commission issued Design Certification of the AP600 standard nuclear reactor design. This culminated an 8-year review of the AP600 design, safety analysis and probabilistic risk assessment. The AP600 is a 600 MWe reactor that utilizes passive safety features that, once actuated, depend only on natural forces such as gravity and natural circulation to perform all required safety functions. These passive safety systems result in increased plant safety and have also significantly simplified plant systems and equipment, resulting in simplified plant operation and maintenance. The AP600 meets NRC deterministic safety criteria and probabilistic risk criteria with large margins. A summary comparison of key passive safety system design features is provided in Table 1. These key features are discussed due to their importance in affecting the key thermal-hydraulic phenomenon exhibited by the passive safety systems in critical areas. The scope of some of the design changes to the AP600 is described. These changes are the ones that are important in evaluating the passive plant design features embodied in the certified AP600 standard plant design. These design changes are incorporated into the AP1000 standard plant design that Westinghouse is certifying under 10 CFR Part 52. In conclusion, this report describes the results of the representative design certification activities that were partially supported by the Nuclear Energy Research Initiative. These activities are unique to AP1000, but are representative of research activities that must be driven to conclusion to realize successful licensing of the next generation of nuclear power plants in the United States

  12. Study of Cost Effective Large Advanced Pressurized Water Reactors that Employ Passive Safety Features

    Energy Technology Data Exchange (ETDEWEB)

    Winters, J. W.; Corletti, M. M.; Hayashi, Y.

    2003-11-12

    A report of DOE sponsored portions of AP1000 Design Certification effort. On December 16, 1999, The United States Nuclear Regulatory Commission issued Design Certification of the AP600 standard nuclear reactor design. This culminated an 8-year review of the AP600 design, safety analysis and probabilistic risk assessment. The AP600 is a 600 MWe reactor that utilizes passive safety features that, once actuated, depend only on natural forces such as gravity and natural circulation to perform all required safety functions. These passive safety systems result in increased plant safety and have also significantly simplified plant systems and equipment, resulting in simplified plant operation and maintenance. The AP600 meets NRC deterministic safety criteria and probabilistic risk criteria with large margins. A summary comparison of key passive safety system design features is provided in Table 1. These key features are discussed due to their importance in affecting the key thermal-hydraulic phenomenon exhibited by the passive safety systems in critical areas. The scope of some of the design changes to the AP600 is described. These changes are the ones that are important in evaluating the passive plant design features embodied in the certified AP600 standard plant design. These design changes are incorporated into the AP1000 standard plant design that Westinghouse is certifying under 10 CFR Part 52. In conclusion, this report describes the results of the representative design certification activities that were partially supported by the Nuclear Energy Research Initiative. These activities are unique to AP1000, but are representative of research activities that must be driven to conclusion to realize successful licensing of the next generation of nuclear power plants in the United States.

  13. Studies on reactor physics

    International Nuclear Information System (INIS)

    Most of the peaceful applications of atomic energy are inherently dependent on advances in the science and technology of nuclear reactors, and aspects of this development are part of a major programme of the International Atomic Energy Agency. The most useful role that the Agency can play is as a co-ordinating body or central forum where the trends can be reviewed and the results assessed. Some of the basic studies are carried out by members of the Agency's own scientific staff. The Agency also convenes groups of experts from different countries to examine a particular problem in detail and make any necessary recommendations. Some of the important subjects are discussed at international scientific meetings held by the Agency. One of the subjects covered by such studies is the physics of nuclear reactors and a specific topic recently discussed was Codes for Reactor Computations, on which a seminar was held in Vienna in April this year. Another The members of the Panel described the development of heavy water reactors, the equipment and methods of research currently used, and plans for further development in their respective countries meeting of Panel of Experts on Heavy Water Lattices was held in Vienna in August 1959

  14. 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.

  15. 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

  16. Studies on influence of sodium void reactivity effect on the concept of the core and safety of advanced fast reactor

    International Nuclear Information System (INIS)

    The paper is devoted to studies on influence of sodium void reactivity effect (SVRE) on safety and technical and economical characteristics of BN-1200 type reactor. Different core options are considered as applied to this reactor. These core options differ in designs, dimensions and, hence, SVRE value. It is shown by the analysis that most flattened core with sodium plenum at the top assures reactor self-protection under beyond design basis accident conditions. Sodium plenum abandonment and core height increase causing SVRE increase deteriorate reactor self-protection, but at the same time, improve some technical and economical characteristics of the reactor. Issues of choosing optimal core design under these conditions are discussed. (author)

  17. 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.

  18. 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.

  19. 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)

  20. 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

  1. 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

  2. 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

  3. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Jorge [Univ. of Utah, Salt Lake City, UT (United States)

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  4. Fusion reactor studies

    International Nuclear Information System (INIS)

    A review is given of fusion reactor systems studies, the objectives of these studies are outlined and some recent conceptual reactor designs are described. The need for further studies in greater depth is indicated so that progress towards a commercial fusion reactor may be consolidated. (U.K.)

  5. 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.

  6. 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.

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. Advanced reactor design and safety objectives - The heavy water reactor perspective

    International Nuclear Information System (INIS)

    This paper provides a summary of the major requirements for future nuclear reactors from CANDU operating station owners based on the various studies and plans prepared. Most of the specific technical requirements for Advanced Heavy Water reactor Systems are based on systematic reviews of current operating CANDU stations to identify opportunities for generic improvements in reliability, operability, maintainability and to address emerging licensing or safety issues. Hence these requirements represent those for the evolutionary development of the Advanced Heavy Water Reactor systems factoring in the considerable operating experience of the CANDU stations. This evolutionary approach to the development of advanced heavy water reactors will be consistent with a philosophy of minimizing the risk to future reactor owners whose requirements are for a reliable, low cost unit

  13. 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)

  14. 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.)

  15. Feasibility Study of Secondary Heat Exchanger Concepts for the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall

    2011-09-01

    The work reported herein represents a significant step in the preliminary design of heat exchanger options (material options, thermal design, selection and evaluation methodology with existing challenges). The primary purpose of this study is to aid in the development and selection of the required heat exchanger for power production using either a subcritical or supercritical Rankine cycle.

  16. Fundamental Studies of the Role of Grain Boundaries on Uniform Corrosion of Advanced Nuclear Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Taheri, Mitra [Drexel Univ., Philadelphia, PA (United States); Motta, Arthur [Pennsylvania State Univ., University Park, PA (United States); Marquis, Emmanuelle [Univ. of Michigan, Ann Arbor, MI (United States)

    2016-05-20

    The main objective of this proposal is to develop fundamental understanding of the role of grain boundaries in stable oxide growth. To understand the process of oxide layer destabilization, it is necessary to observe the early stages of corrosion. During conventional studies in which a sample is exposed and examined after removal from the autoclave, the destabilization process will normally have already taken place, and is only examined post facto. To capture the instants of oxide destabilization, it is necessary to observe it in-situ; however significant question always arise as to the influence of the corrosion geometry and conditions on the corrosion process. Thus a combination of post facto examinations and in-situ studies is proposed, which also combines state-of-the-art characterization techniques to derive a complete understanding of the destabilization process and the role of grain boundaries.

  17. Methodology for studies of natural circulation in advanced reactors using CFD

    International Nuclear Information System (INIS)

    In the area of nuclear engineering, some problems of fluid mechanics and heat transfer are frequently present, so we always have to get better analytical and experimental methods for solving these specific problems. It is presented an analysis of stability of the phenomenon of natural circulation for single-phase flow in a closed loop. This paper shows a stabilized finite element formulation developed for the solution of the Navier-Stokes and energy equations in cartesian coordinates. When working with computer programs, an important step to do is to validate the code developed. So, for the validation of our computational code, the study the natural circulation phenomenon in a thermal hydraulic loop was performed. It was tested the computational model by comparing the values of the mass flow along the time with results obtained with other numerical modelling. (authors)

  18. Additional comments on CRP 'Studies of advanced reactor technology options for effective incineration of radioactive waste'

    International Nuclear Information System (INIS)

    The CRP should represent a 'leap forward' in understanding of some areas in transmuter physics and safety: concentrate on transmuters; critical and subcritical systems; cover neutronic, thermal-hydraulic/mechanical and fuel issues; concentrate on issues of transient phenomena (dynamics). The CRP should have the freedom to deal with innovative approaches and ideas. Investigation of transient states should cover the design basis condition and design extension condition area: Identify where information if lacking; identify where information may be needed in the future; provide information if possible. Example of design issues to be investigated are shown with examples of neutronics and thermal-hydraulic/mechanical issues to be studied. An example of fuel behaviour under transient conditions to be investigated is described

  19. AB INITIO STUDY OF ADVANCED METALLIC NUCLEAR FUELS FOR FAST BREEDER REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Landa, A; Soderlind, P; Grabowski, B; Turchi, P A; Ruban, A V; Vitos, L

    2012-04-23

    Density-functional formalism is applied to study the ground state properties of {gamma}-U-Zr and {gamma}-U-Mo solid solutions. Calculated heats of formation are compared with CALPHAD assessments. We discuss how the heat of formation in both alloys correlates with the charge transfer between the alloy components. The decomposition curves for {gamma}-based U-Zr and U-Mo solid solutions are derived from Ising-type Monte Carlo simulations. We explore the idea of stabilization of the {delta}-UZr{sub 2} compound against the {alpha}-Zr (hcp) structure due to increase of Zr d-band occupancy by the addition of U to Zr. We discuss how the specific behavior of the electronic density of states in the vicinity of the Fermi level promotes the stabilization of the U{sub 2}Mo compound. The mechanism of possible Am redistribution in the U-Zr and U-Mo fuels is also discussed.

  20. 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)

  1. 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

  2. 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

  3. 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.)

  4. 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.)

  5. 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)

  6. 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)

  7. 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

  8. 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

  9. Comparative study on level-2 PSA modeling for an advanced pressurized water reactor during basic design phase in Korea

    International Nuclear Information System (INIS)

    Concerning the probabilistic safety assessment (PSA) for the containment response and the source term characteristics under hypothetical core damage accidents, during the basic design phase of an advanced pressurized water reactor (APWR), a level-2 PSA modeling was developed. In the present paper, a comparative study on the modeling was performed to obtain insights related to safety adequacy and vulnerability of the APWR basic design with some severe accident prevention/mitigation features (structures, systems, and components). The features consist of containment structure based on enhanced design pressure, improved cavity configuration, hydrogen mitigation system, cavity flooding system with both active and passive spillways, emergency containment spray backup system, and safety depressurization and vent system. The level-2 PSA modeling was implemented by reflecting limited results obtained from the level-1 PSA, the deterministic severe accident analysis, and the deterministic containment ultimate pressure capacity analysis using the above features. In spite of the limitation of the level-2 PSA modeling during the basic design phase, it is found that the quantitative safety of the APWR basic design was enhanced to the extent that containment failure frequency (CFF) and conditional containment failure probability (CCFP) meet CFF goal and containment performance goal, respectively. Additionally, compared with the level-2 PSA modeling results of the APWR and those of other plants, the qualitative level of safety was identified with respect to the basic design aspects vulnerable to source term release sequences. Finally, to reduce severe accident vulnerability, it was recommended that actions for lowering containment heat removal loss leading to the late failure be developed and effort for reducing the direct release of fission product due to the SGTR-induced containment bypass be made

  10. 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

  11. Advanced Reactors Thermal Energy Transport for Process Industries

    Energy Technology Data Exchange (ETDEWEB)

    P. Sabharwall; S.J. Yoon; M.G. McKellar; C. Stoots; George Griffith

    2014-07-01

    The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as liquid fuel production, district heating, desalination, hydrogen production, and other process heat applications, etc. Some of the major technology challenges that must be overcome before the advanced reactors could be licensed on the reactor side are qualification of next generation of nuclear fuel, materials that can withstand higher temperature, improvement in power cycle thermal efficiency by going to combined cycles, SCO2 cycles, successful demonstration of advanced compact heat exchangers in the prototypical conditions, and from the process side application the challenge is to transport the thermal energy from the reactor to the process plant with maximum efficiency (i.e., with minimum temperature drop). The main focus of this study is on doing a parametric study of efficient heat transport system, with different coolants (mainly, water, He, and molten salts) to determine maximum possible distance that can be achieved.

  12. Studies on influence of sodium void reactivity effect on the concept of the core and safety of advanced fast reactor

    International Nuclear Information System (INIS)

    The paper is devoted to studies on the influence of the sodium void reactivity effect (SVRE) on the safety and technical and economical characteristics of the BN-1200-type reactor. Different core options are considered for application to this reactor. These core options differ in design, dimensions, and, hence, SVRE value. It is shown by the analysis that the most flattened core with sodium plenum at the top assures reactor self-protection under beyond-design-basis accident conditions. Sodium plenum abandonment and core height increase causing an SVRE value increase deteriorate reactor self-protection, but at the same time, improve some technical and economical characteristics of the reactor. Self-protection means the possibility to avoid rapid core meltdown under conditions of the above-listed beyond-design accidents. The possibility of controlling beyond-design accidents (for instance, by restoring the power supply of the main pumps in a rather short time) is taken into account. Issues of choosing the optimal core design under these conditions are discussed. (author)

  13. 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)

  14. Reactor scram events in the updated PIUS 600 advanced reactor design

    International Nuclear Information System (INIS)

    The PIUS advanced reactor is a 640-MWe pressurized water reactor concept developed by Asea Brown Boveri. A unique feature of PIUS is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. Los Alamos supported the US Nuclear Regulatory Commission's preapplication review of the PIUS reactor. Baseline calculations of the PIUS design were performed for active and passive reactor scrams using TRAC-PF1/MOD2. Additional sensitivity studies examined flow blockage and boron dilution events to explore the robustness of the PIUS concept for low-probability combination events following active-system scrams

  15. 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

  16. Analytical and Experimental Study of The Effects of Non-Condensable in a Passive Condenser System for The Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    The main goal of the project is to study analytically and experimentally condensation heat transfer for the passive condenser system relevant to the safety of next generation nuclear reactor such as Simplified Boiling Water Reactor (BWR). The objectives of this three-year research project are to: (1) obtain experimental data on the phenomenon of condensation of steam in a vertical tube in the presence of non-condensable for flow conditions of PCCS, (2) develop a analytic model for the condensation phenomena in the presence of non-condensable gas for the vertical tube, and (3) assess the RELAP5 computer code against the experimental data. The project involves experiment, theoretical modeling and a thermal-hydraulic code assessment. It involves graduate and undergraduate students' participation providing them with exposure and training in advanced reactor concepts and safety systems

  17. Analytical and Experimental Study of The Effects of Non-Condensable in a Passive Condenser System for The Advanced Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shripad T. Revankar; Seungmin Oh

    2003-09-30

    The main goal of the project is to study analytically and experimentally condensation heat transfer for the passive condenser system relevant to the safety of next generation nuclear reactor such as Simplified Boiling Water Reactor (BWR). The objectives of this three-year research project are to: (1) obtain experimental data on the phenomenon of condensation of steam in a vertical tube in the presence of non-condensable for flow conditions of PCCS, (2) develop a analytic model for the condensation phenomena in the presence of non-condensable gas for the vertical tube, and (3) assess the RELAP5 computer code against the experimental data. The project involves experiment, theoretical modeling and a thermal-hydraulic code assessment. It involves graduate and undergraduate students' participation providing them with exposure and training in advanced reactor concepts and safety systems

  18. 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

  19. Advancing liquid metal reactor technology with nitride fuels

    International Nuclear Information System (INIS)

    A review of the use of nitride fuels in liquid metal fast reactors is presented. Past studies indicate that both uranium nitride and uranium/plutonium nitride possess characteristics that may offer enhanced performance, particularly in the area of passive safety. To further quantify these effects, the analysis of a mixed-nitride fuel system utilizing the geometry and power level of the US Advanced Liquid Metal Reactor as a reference is described. 18 refs., 2 figs., 2 tabs

  20. 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

  1. 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.

  2. 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)

  3. 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

  4. 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

  5. 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

  6. Thermohydraulic study of the reflooding phase after a loss of coolant accident in an advanced pressurized water reactor

    International Nuclear Information System (INIS)

    The aim of this work is to analyse the behaviour of some different Advanced Pressurized Water Reactor (APWR) designs during the reflooding phase following a large break (double ended) loss of coolant accident (LOCA). The FLUT code and its most important physical models and correlations will be shortly discussed. The major changes will be described addressing the reasons for their introduction. A special attention is given to the description of a newly developed droplet-model, which allows a better modelling of the progression of the quench front. In order to assess the updated correlations and models, many forced reflooding experiments were evaluated. The results of these post-test calculations will be shown. The calculations for two APWR designs will be then compared with that for the reference PWR, pointing out the most important differences. (orig.)

  7. 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

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  8. 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.

  9. 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

  10. 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.

  11. 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.

  12. 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)

  13. 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.

  14. 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

  15. Neutronic challenges of advanced boiling water reactor designs

    International Nuclear Information System (INIS)

    The advancement of Boiling Water Reactor technology has been under investigation at the Center for Advance Nuclear Energy Systems at MIT. The advanced concepts under study provide economic incentives through enabling further power uprates (i.e. increasing vessel power density) or better fuel cycle uranium utilization. The challenges in modeling of three advanced concepts with focus on neutronics are presented. First, the Helical Cruciform Fuel rod has been used in some Russian reactors, and studied at MIT for uprating the power in LWRs through increased heat transfer area per unit core volume. The HCF design requires high fidelity 3D tools to assess its reactor physics behavior as well as thermal and fuel performance. Second, an advanced core design, the BWR-HD, was found to promise 65% higher power density over existing BWRs, while using current licensing tools and existing technology. Its larger assembly size requires stronger coupling between neutronics and thermal hydraulics compared to the current practice. Third is the reduced moderation BWRs, which had been proposed in Japan to enable breeding and burning of fuel as an alternative to sodium fast reactors. Such technology suffers from stronger sensitivity of its neutronics to the void fraction than the traditional BWRs, thus requiring exact modeling of the core conditions such as bypass voiding, to correctly characterize its performance. (author)

  16. Astronaut Kevin Chilton works with advanced cell reactor

    Science.gov (United States)

    1994-01-01

    Astronaut Kevin P. Chilton, pilot, works with an advanced cell reactor, which incorporated the first ever videomicroscope, on the Space Tissue Loss (STL-B) experiment on the Space Shuttle Endeavour's middeck. This experiment studied cell growth during the STS-59 mission.

  17. 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)

  18. 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

  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. 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)

  1. 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.

  2. 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

  3. 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)

  4. 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

  5. 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

  6. Mirror reactor surface study

    International Nuclear Information System (INIS)

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included

  7. 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

  8. 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.

  9. Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

    International Nuclear Information System (INIS)

    This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A ampersand 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met

  10. Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

    Energy Technology Data Exchange (ETDEWEB)

    1994-01-15

    This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A & 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met.

  11. 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)

  12. Developments of new advanced techniques for use of a reactor in livestock studies and expansion of their utilization

    International Nuclear Information System (INIS)

    There remain many problems to be solved on the absorption, metabolisms, accumulations and functions of minor elements in domestic animals. Development of new techniques using reactor was attempted to produce more domestic animals and increase livestock products by resolving such problems. Prompt γ-ray analysis (PGA) was used to analyze biological samples in this study. Water added with sodium borate (100 ppm) was given to the sheep and the amounts of boron in their feces were determined by PGA. The amounts in the feces increased from the first day of the administration of boron water and reached the peak on the 8th day. The concentration was ranging in 34-340 ppm, whereas that of the control was in 34-40 ppm. Then, to examine the applicability of activable tracers; samarium (Sm), ytterbium (Yb) and lanthanum (La) as a marker, the food passing time through the digestive track was determined for calves according to activation analysis. Neutron irradiation of the feces sample from the calves was conducted by reactor JRR-3M. The measuring sensitivity was 0.1-1.0 μg for either of these elements. (M.N.)

  13. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  14. Study of future reactors

    International Nuclear Information System (INIS)

    Today, more than 420 large reactors with a gross output of close to 350 GWe supply 20 percent of world electricity needs, accounting for less than 5 percent of primary energy consumption. These figures are not expected to change in the near future, due to suspended reactor construction in many countries. Nevertheless, world energy needs continue to grow: the planet's population already exceeds five billion and is forecast to reach ten billion by the middle of the next century. Most less developed countries have a very low rate of energy consumption and, even though some savings can be made in industrialized countries, it will become increasingly difficult to satisfy needs using fossil fuels only. Furthermore, there has been no recent breakthrough in the energy landscape. The physical feasibility of the other great hope of nuclear energy, fusion, has yet to be proved; once this has been done, it will be necessary to solve technological problems and to assess economic viability. Although it is more ever necessary to pursue fusion programs, there is little likelihood of industrial applications being achieved in the coming decades. Coal and fission are the only ways to produce massive amounts of energy for the next century. Coal must overcome the pollution problems inherent in its use; fission nuclear power has to gain better public acceptance, which is obviously colored by safety and waste concerns. Most existing reactors were commissioned in the 1970s; reactor lifetime is a parameter that has not been clearly established. It will certainly be possible to refurbish some to extend their operation beyond the initial target of 30 or 40 years. But normal advances in technology and safety requirements will make the operation of the oldest reactors increasingly difficult. It becomes necessary to develop new generations of nuclear reactors, both to replace older ones and to revive plant construction in their countries that are not yet equipped or that have halted their

  15. 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

  16. The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D3He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

  17. The ARIES tokamak reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  18. 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

  19. 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)

  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. 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)

  2. 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.

  3. 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

  4. Optimizing advanced liquid metal reactors for burning actinides

    International Nuclear Information System (INIS)

    In this report, the process to design an Advanced Liquid Metal Reactor (ALMR) for burning the transuranic part of nuclear waste is discussed. The influence of design parameters on ALMR burner performance is studied and the results are incorporated in a design schedule for optimizing ALMRs for burning transuranics. This schedule is used to design a metallic and an oxide fueled ALMR burner to burn as much as possible transurancis. The two designs burn equally well. (orig.)

  5. 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

  6. 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

  7. 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)

  8. 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)

  9. 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.

  10. 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)

  11. Advanced reactor design and safety objectives. The heavy water reactor perspective

    International Nuclear Information System (INIS)

    The development objectives of advanced heavy water reactors (AHWRs) should be guided by the requirements of the operating utilities. The paper provides a summary of the major requirements for future nuclear reactors from CANDU operating station owners based on the various studies and plans prepared. Most of the specific technical requirements for AHWR systems are based on systematic reviews of current operating CANDU stations. Hence these requirements represent those for the evolutionary development of AHWR systems, factoring in the considerable operating experiences of the CANDU stations. The requirements for the new HWR designs can be summarized under economic objectives, safety objectives, operational objectives and other utility requirements. 2 refs

  12. 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

  13. 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

  14. 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

  15. 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

  16. Using Advanced Fuel Bundles in CANDU Reactors

    International Nuclear Information System (INIS)

    Improving the exit fuel burnup in CANDU reactors was a long-time challenge for both bundle designers and performance analysts. Therefore, the 43-element design together with several fuel compositions was studied, in the aim of assessing new reliable, economic and proliferation-resistant solutions. Recovered Uranium (RU) fuel is intended to be used in CANDU reactors, given the important amount of slightly enriched Uranium (~0.96% w/o U235) that might be provided by the spent LWR fuel recovery plants. Though this fuel has a far too small U235 enrichment to be used in LWR's, it can be still used to fuel CANDU reactors. Plutonium based mixtures are also considered, with both natural and depleted Uranium, either for peacefully using the military grade dispositioned Plutonium or for better using Plutonium from LWR reprocessing plants. The proposed Thorium-LEU mixtures are intended to reduce the Uranium consumption per produced MW. The positive void reactivity is a major concern of any CANDU safety assessment, therefore reducing it was also a task for the present analysis. Using the 43-element bundle with a certain amount of burnable poison (e.g. Dysprosium) dissolved in the 8 innermost elements may lead to significantly reducing the void reactivity. The expected outcomes of these design improvements are: higher exit burnup, smooth/uniform radial bundle power distribution and reduced void reactivity. Since the improved fuel bundles are intended to be loaded in existing CANDU reactors, we found interesting to estimate the local reactivity effects of a mechanical control absorber (MCA) on the surrounding fuel cells. Cell parameters and neutron flux distributions, as well as macroscopic cross-sections were estimated using the transport code DRAGON and a 172-group updated nuclear data library. (author)

  17. 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.

  18. 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

  19. 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

  20. A methodological study on organizing an intelligent CAD/CAE system for conceptual design of advanced nuclear reactor system

    International Nuclear Information System (INIS)

    In order to shorten the time span of design work and enhance both consistency and rationality of design products, the authors are now investigating an intelligent CAD/CAE system to support cooperative works by many specialists by adopting object-oriented approach. In this paper, the cognitive aspect of design activities of specialists in the conceptual design phase of nuclear reactors is discussed. The activities of the specialists in their design analysis process are highly knowledge-based and goal-oriented. The characteristics of the activities are 1) hierarchization of design goal into sub-goals, 2) prioritization of design sub-goals and step-by-step practise of design analysis, and 3) abstraction of real-world space structure into more simplified space structure to cope with theoretical treatment. Based on these consideration, a conceptual design model of specialists' activities composed of attribute modeling and design expertise knowledge base is proposed. The 'principle of functional independence' proposed by Sue is applied to bridge between the attribute modeling and design expertise knowledge base. The intelligent CAD/CAE system is now under development by focusing on the conceptual design of a space power reactor core utilizing thermo-ionic fuel elements as direct thermo-to-electric conversion. A program to calculate thermo-hydraulics of reactor core and thermo-ionic power generation has been developed. An interface has been also developed in order to communicate with the specialists at JAERI by E-mail concerning the interactive calculation between our calculation and the neutronics calculation of reactor core. (orig.)

  1. 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

  2. Reactor fault simulation at the closure of the Windscale advanced gas-cooled reactor: analysis of reactor transient tests

    International Nuclear Information System (INIS)

    The testing of fault transient analysis methods by direct simulation of fault sequences on a commercial reactor is clearly excluded on safety and economic grounds. The closure of the Windscale prototype advanced gas-cooled reactor (WAGR) therefore offered a unique opportunity to test fault study methods under extreme conditions relatively unfettered by economic constraints, although subject to appropriate safety regulations. One aspect of these important experiments was a series of reactor transient tests. The objective of these reactor transients was to increase confidence in the fault study computer models used for commercial AGR safety assessment by extending their range of validation to cover large amplitude and fast transients in temperature, power and flow, relevant to CAGR faults, and well beyond the conditions achievable experimentally on commercial reactors. A large number of tests have now been simulated with the fault study code KINAGRAX. Agreement with measurement is very good and sensitivity studies show that such discrepancies as exist may be due largely to input data errors. It is concluded that KINAGRAX is able to predict steady state conditions and transient amplitudes in both power and temperature to within a few percent. (author)

  3. 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

  4. 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

  5. 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.

  6. 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)

  7. Recycled uranium: An advanced fuel for CANDU reactors

    International Nuclear Information System (INIS)

    The use of recycled uranium (RU) fuel offers significant benefits to CANDU reactor operators particularly if used in conjunction with advanced fuel bundle designs that have enhanced performance characteristics. Furthermore, these benefits can be realised using existing fuel production technologies and practices and with almost negligible change to fuel receipt and handling procedures at the reactor. The paper will demonstrate that the supply of RU as a ceramic-grade UO2 powder will increasingly become available as a secure option to virgin natural uranium and slightly enriched uranium(SEU). In the context of RU use in Canadian CANDU reactors, existing national and international transport regulations and arrangements adequately allow all material movements between the reprocessor, RU powder supplier, Canadian CANDU fuel manufacturer and Canadian CANDU reactor operator. Studies have been undertaken of the impact on personnel dose during fuel manufacturing operations from the increased specific activity of the RU compared to natural uranium. These studies have shown that this impact can be readily minimised without significant cost penalty to the acceptable levels recognised in modem standards for fuel manufacturing operations. The successful and extensive use of RU, arising from spent Magnox fuel, in British Energy's Advanced Gas-Cooled reactors is cited as relevant practical commercial scale experience. The CANFLEX fuel bundle design has been developed by AECL (Canada) and KAERI (Korea) to facilitate the achievement of higher bum-ups and greater fuel performance margins necessary if the full economic potential of advanced CANDU fuel cycles are to be achieved. The manufacture of a CANFLEX fuel bundle containing RU pellets derived from irradiated PWR fuel reprocessed in the THORP plant of BNFL is described. This provided a very practical verification of dose modelling calculations and also demonstrated that the increase of external activity is unlikely to require any

  8. 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

  9. 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

  10. 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.)

  11. 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

  12. 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

  13. 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.

  14. Effect of horizontal cross flow on the heat transfer form the moderator bricks in the Advanced Gas-cooled Reactor: A CFD study

    International Nuclear Information System (INIS)

    Highlights: • Small cross flow can have a significant effect on the cooling of the graphite brick of UK Advanced Gas-cooled Reactor (AGR). • Cross flow causes the peak temperature of the brick to reduce. • Cross flow causes a redistribution of the main axial downward flow. • Cross flow enhances the heat transfer in some of passages of small clearances of the key/keyway connections. -- Abstract: The paper reports an investigation of the effect of the horizontal cross flow on the temperature of the moderator brick in the UK Advanced Gas-cooled Reactor (AGR) using computational fluid dynamics (CFD) with a conjugate heat transfer model for the solid and fluid. The commercial software package ANSYS Fluent is used to study the correlation between the flow structure and heat transfer which ultimately determine the temperature within the system. The CFD model comprises the full axial length of one-half of a typical fuel channel (assuming symmetry) and part of the neighbouring channels on either side. Two sets of simulations have been carried out, namely, one with cross flow and one without cross flow. The effect of cross flow has subsequently been derived by comparing the results from these two groups of simulations. The study shows that a small cross flow can have a significant effect on the cooling of the graphite brick, causing the peak temperature of the brick to reduce significantly. Two mechanisms are identified to be responsible for this. Firstly, the small cross flow causes a significant redistribution of the main axial downward flow and this leads to an enhancement of heat transfer in some of the small clearances, and an impairment in others although overall, the enhancement is dominant leading to a better cooling. Secondly, the cross flow makes effective use of the small clearances between the key/keyway connections which increases the effective heat transfer area, hence increasing the cooling. Under the conditions of no cross flow, these areas remain

  15. 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)

  16. A neutronic study on advanced sodium cooled fast reactor cores with thorium blankets for effective burning of transuranic nuclides

    International Nuclear Information System (INIS)

    Highlights: • SFR burner core configurations are explored and analyzed for effective use of thorium blankets. • Thorium blankets can significantly improve SFR burner core performances. • No recycling or partial recycling of Th blankets with multi-batches is very effective. - Abstract: In this paper, new design concepts of sodium cooled fast reactor (SFR) cores having thorium blanket are suggested for pursuing effective burning of TRU (transuranics) nuclides from LWR spent fuels and their neutronic performances are analyzed. Several core configurations having different arrangements of thorium blankets are explored to improve the core performances and safety-related parameters including sodium void worth which is one of main concerns on safety of SFR cores. Specifically, axial and radial thorium blankets are considered for two type cores. The first one is the typical annular type cores having two different fuel regions where axial thorium blankets are placed in the axially central regions while the second one is the single fuel region cores having central non-fuel region where the axial blanket and radial blankets are considered. Also, the effects of the recycling options and fuel management schemes of the used thorium blanket on the core performances are analyzed. The core performance analyses show that thorium blankets with no recycling option and multi-batch fuel management schemes are very effective to improve the core performances including burnup reactivity swing, sodium void worth and TRU consumption rate

  17. Digital control application for the advanced boiling water reactor

    International Nuclear Information System (INIS)

    The Advanced Boiling Water Reactor (ABWR) is a 1300 MWe class Nuclear Power Plant whose design studies and demonstration tests are being performed by the three manufacturers, General Electric, Toshiba and Hitachi, under requirement specifications from the Tokyo Electric Power Company. The goals are to apply new technology to the BWR in order to achieve enhanced operational efficiencies, improved safety measures and cost reductions. In the plant instrumentation and control areas, traditional analog control equipment and wire cables will be replaced by distributed digital microprocessor based control units communicating with each other and the control room over fiber optic multiplexed data buses

  18. Novel Engineered Refractory Materials for Advanced Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Shannon, Steven [North Carolina State Univ., Raleigh, NC (United States); Eapen, Jacob [North Carolina State Univ., Raleigh, NC (United States); Maria, Jon-Paul [North Carolina State Univ., Raleigh, NC (United States); Weber, William [Univ. of Tennessee, Knoxville, TN (United States)

    2015-05-01

    This report summarizes the results of DOE-NEUP grant 10-853 titled “Novel engineered refractory materials for advanced reactor applications”. The project spanned 48 months (36 months under the original grant plus a 12 month no cost extension). The overarching goal of this work was to fabricate and characterize refractory materials engineered at the atomic scale with emphasis on their tolerance to accumulated radiation damage. With an emphasis on nano-scale structure, this work included atomic scale simulation to study the underlying mechanisms for modified radiation tolerance at these atomic scales.

  19. 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

  20. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  1. 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.

  2. 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

  3. Advanced sodium fast reactor unit concept

    International Nuclear Information System (INIS)

    The paper presents status of development for 1200 MW power unit with sodium fast reactor for commercial construction in the Russian Federation. General characteristics of the reactor plant (RP) and power unit as well as goals that shall be achieved because of design development are described. The power unit design is based on technical decisions, which have been partially proven during sodium reactor operation in Russia and partially have been validated by R and D work for BN-800 RP. At the same time, new technical decisions are applied that improve safety and technical-and-economic indices. To validate them, the corresponding R and D work shall be performed. It is planned to construct the pilot power unit in 2020 and to put into operation the next commercial power units of this type using plutonium generated in the thermal reactors. (author)

  4. An advanced method of heterogeneous reactor theory

    International Nuclear Information System (INIS)

    Recent approaches to heterogeneous reactor theory for numerical applications were presented in the course of 8 lectures given in JAERI. The limitations of initial theory known after the First Conference on Peacefull Uses of Atomic Energy held in Geneva in 1955 as Galanine-Feinberg heterogeneous theory:-matrix from of equations, -lack of consistent theory for heterogeneous parameters for reactor cell, -were overcome by a transformation of heterogeneous reactor equations to a difference form and by a development of a consistent theory for the characteristics of a reactor cell based on detailed space-energy calculations. General few group (G-number of groups) heterogeneous reactor equations in dipole approximation are formulated with the extension of two-dimensional problem to three-dimensions by finite Furie expansion of axial dependence of neutron fluxes. A transformation of initial matrix reactor equations to a difference form is presented. The methods for calculation of heterogeneous reactor cell characteristics giving the relation between vector-flux and vector-current on a cell boundary are based on a set of detailed space-energy neutron flux distribution calculations with zero current across cell boundary and G calculations with linearly independent currents across the cell boundary. The equations for reaction rate matrices are formulated. Specific methods were developed for description of neutron migration in axial and radial directions. The methods for resonance level's approach for numerous high-energy resonances. On the basis of these approaches the theory, methods and computer codes were developed for 3D space-time react or problems including simulation of slow processes with fuel burn-up, control rod movements, Xe poisoning and fast transients depending on prompt and delayed neutrons. As a result reactors with several thousands of channels having non-uniform axial structure can be feasibly treated. (author)

  5. Advanced Plasma Pyrolysis Assembly (PPA) Reactor and Process Development

    Science.gov (United States)

    Wheeler, Richard R., Jr.; Hadley, Neal M.; Dahl, Roger W.; Abney, Morgan B.; Greenwood, Zachary; Miller, Lee; Medlen, Amber

    2012-01-01

    Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA's Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development.

  6. Loss of feed water analyses of advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    The proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. Passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power level with no primary coolant pumps. The case analysed in this paper is the loss of feedwater to steam drum which results in decrease in heat removal from core. This also causes increase in reactor pressure. Further consequences depend upon various protective and engineered safeguard systems like relief system, reactor trip, isolation condenser and advanced accumulator. Analysis has been done using code RELAP5/MOD3.2. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, qualities and flow in different part of Primary Heat Transport (PHT) system. (author)

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

    International Nuclear Information System (INIS)

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

  8. Advanced High Temperature Reactor Systems and Economic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Qualls, A L [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience

  9. Advanced High Temperature Reactor Systems and Economic Analysis

    International Nuclear Information System (INIS)

    The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience with

  10. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

    1999-04-01

    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

  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. 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)

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

    International Nuclear Information System (INIS)

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

  14. Chernobyl reactor transient simulation study

    International Nuclear Information System (INIS)

    This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

  15. Studies on nuclear reactor design

    International Nuclear Information System (INIS)

    this thesis presents two studies for safety aspects in nuclear reactor design. the fission process that occurs in the reactor core is the most important process for the harmful effect of produced radiation especially neutrons with different energies and gamma radiations for their strong penetrability . so studying the criticality of the fissile materials in the reactor is one of the most important safety aspects for the reactor design, the attenuation of the neutrons and gammas using suitable shielding materials with suitable thicknesses is the second study that is discussed in this thesis

  16. 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)

  17. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts

  18. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  19. Experimental studies on interactions of molten LiF-NaF salt with some candidate structural materials for components of advanced nuclear reactors

    International Nuclear Information System (INIS)

    Interaction of molten 60 LiF - 40 NaF (% mol) salt with candidate structural materials for components of advanced nuclear reactors has been studied using electron probe microanalysis (EPMA) and inductively coupled plasma atomic emission spectrometry (ICP-OES). The corrosion of structural materials (stainless steel, Ni base alloy, nickel), which was induced by the molten salt melt, has been examined in dependence on the time of exposure at operating temperature of 680 deg C. The above choice of the two analytical techniques made it possible to assess on the whole the extent of corrosion. The corrosion phenomena in structural materials were investigated using EPMA. Corrosion-released elements dissolved in solidified molten salt were determined after salt dissolution by means of ICP-OES. The LiF-NaF melt produced corrosion, which proved as a surficial modification of a structural material and a trace contamination of the melt itself. The X-ray maps by EPMA with its 1-μm lateral resolution revealed compositional changes in structural materials, such as, e. g. regular depletion of Cr in alloy A686 to the depth of 10 - 25 μm. While the lateral resolution of LA-ICP-MS with the applied laser spot diameter of 25 μm was not exactly adequate to mapping of the corroded material section and, consequently, yielded less information in comparison with EPMA, this technique was quite sufficient for the mapping of elemental content changes in solidified salt profile. Finally, nickel was proved to be the most resistant material. It was concluded that: (i) EPMA study, involving semi-quantitative elemental mapping / content profiling and detailed spot quantitative analyses makes it possible to obtain quantitative assessment of the corrosion process; (ii) qualitative profiles are provided by LA-ICP-MS, which needs further development on quantification procedure based on matched calibration samples. (author)

  20. 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)

  1. Mirror advanced reactor study (MARS)

    International Nuclear Information System (INIS)

    The agenda for the meeting is as follows: (1) basic Tandem Mirror approach, (2) baseline design, (3) transition and Yin-Yang coils, (4) drift pump physics, (5) drift pump coil, (6) Fokker-Planck analysis, (7) ignition-alpha pumping, (8) neutral beam status, (9) axicell layout, (10) axicell radiation levels, (11) ICRH system, (12) central cell cost optimization, (13) central cell coil design, (14) gridless direct converter, (15) direct converter directions, (16) end cell structure, (17) corrosion-double wall HX, (18) central cell maintenance, (19) radioactivity, (20) PbLi blanket design, and (21) MARS schedule

  2. Models for transient analyses in advanced test reactors

    OpenAIRE

    Gabrielli, Fabrizio

    2011-01-01

    Several strategies are developed worldwide to respond to the world’s increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe whi...

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

    International Nuclear Information System (INIS)

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

  4. Advanced Multiphysics Modeling of Fast Reactor Fuel Behavior

    International Nuclear Information System (INIS)

    Evaluation of fast reactor fuel thermo-mechanical performance using fuel performance codes is a key aspect of advanced fast reactors designs. Those fuel performance codes capture the multiphysics nature of fuel behavior during irradiation where different, mostly interdependent, phenomena are taking place. Existing fuel performance codes do not fully capture those interdependencies and present the different phenomena through de-coupled models. Recent developments in multiphysics simulation capabilities and availability of advanced computing platforms led to advancements in simulation of nuclear fuel behavior. This paper presents current experiences in applying different multiphysics simulation platforms to evaluation of fast reactors metallic fuel behavior. Full 3D finite element simulation platforms that include capabilities to fully couple key fuel behavior models are discussed. Issues associated with coupling metallic fuels phenomena, such as fission gas models and constituent distribution models, with thermo-mechanical finite element platforms, as well as different coupling schemes are also discussed. (author)

  5. Actinide transmutation in the advanced liquid metal reactor (ALMR)

    International Nuclear Information System (INIS)

    The Advanced Liquid Metal Reactor (ALMR) is a US Department of Energy (DOE) sponsored fast reactor design based on the Power Reactor, Innovative Small Module (PRISM) concept originated by General Electric. The current reference design is a 471 MWt modular reactor loaded with ternary metal fuel. This paper discusses actinide transmutation core designs that fit the design envelope of the ALMR and utilize spent LWR fuel as startup material and makeup. Actinide transmutation may be accomplished in the ALMR by using either a breeding or burning configuration. Lifetime actinide mass consumption is calculated as well as changes in consumption behaviour throughout the lifetime of the reactor. Impacts on system operational and safety performance are evaluated in a preliminary fashion. (author). 3 refs, 6 figs, 3 tabs

  6. 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

  7. Perspective of nuclear energy and advanced reactors

    International Nuclear Information System (INIS)

    Future nuclear energy growth will be the result of the contributions of every single plant being constructed or projected at present as it is connected to the grid. As per IAEA, there exists presently 34 nuclear power plants under construction 81 with the necessary permits and funding and 223 proposed, which are plants seriously pursuing permits and financing. This means that in a few decades the number of nuclear power plants in operation will have doubled. This growth rate is characterised by the incorporation of new countries to the nuclear club and the gradually increasing importance of Asian countries. During this expansive phase, generation III and III+designs are or will be used. These designs incorporate the experience from operating plants, and introduce innovations on rationalization design efficiency and safety, with emphasis on passive safety features. In a posterior phase, generation IV designs, presently under development, will be employed. Generation IV consists of several types of reactors (fast reactors, very high temperature reactors, etc), which will improve further sustain ability, economy, safety and reliability concepts. The described situation seems to lead to a renaissance of the nuclear energy to levels hardly thinkable several years ago. (Author)

  8. Neutrino Oscillation Studies with Reactors

    CERN Document Server

    Vogel, Petr; Zhang, Chao

    2015-01-01

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  9. Fusion reactor design towards radwaste minimum with advanced shield material

    International Nuclear Information System (INIS)

    A new concept of fusion reactor design is proposed to minimize the radioactive waste of the reactor. The main point of the concept is to clear massive structural components located outside the neutron shield from regulatory control. The concept requires some reinforcement of shielding with an advanced shield material such as a metal hydride, detriation, and tailoring of a detrimental element from the superconductor. Our assessment confirmed a large impact of the concept on radwaste reduction, in that it reduces the radwaste fraction of a fusion reactor A-SSTR2 from 92 wt.% to 17 wt.%. (author)

  10. Spectrophotometric Procedure for Fast Reactor Advanced Coolant Manufacture Control

    Science.gov (United States)

    Andrienko, O. S.; Egorov, N. B.; Zherin, I. I.; Indyk, D. V.

    2016-01-01

    The paper describes a spectrophotometric procedure for fast reactor advanced coolant manufacture control. The molar absorption coefficient of dimethyllead dibromide with dithizone was defined as equal to 68864 ± 795 l·mole-1·cm-1, limit of detection as equal to 0.583 · 10-6 g/ml. The spectrophotometric procedure application range was found to be equal to 37.88 - 196.3 g. of dimethyllead dibromide in the sample. The procedure was used within the framework of the development of the method of synthesis of the advanced coolant for fast reactors.

  11. 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

  12. Sloshing and fluid-structure interaction in a 400-MWe pool-type advanced fast reactor

    International Nuclear Information System (INIS)

    This paper describes the seismic analysis of a 400-MWe advanced fast reactor under 0.3 g SSE ground excitation. Two types of analyses are performed - the sloshing analysis and the fluid-structure interaction analysis. In the sloshing analysis, the sloshing frequency and wave patterns are calculated. The maximum wave height and the sloshing forces exerted on the submerged components and the primary tank are evaluated. In the fluid-structure interaction analysis, the maximum horizontal acceleration for the reactor core and the relative displacement between the reactor core and UIS are examined. The fluid-coupling phenomena between various components are investigated. Seismic stresses at critical areas are examined. The results obtained from this study are very useful to the design of the advanced reactors. Meanwhile, the computer code FLUSTR-ANL has proved to be a useful analytical tool for assessing the complicated seismic fluid-structure interactions and sloshing in the fast reactor systems. 10 refs., 25 figs

  13. A general description of AARE [Advanced Analysis for Reactor Engineering]: A modular system for advanced analysis of reactor engineering

    International Nuclear Information System (INIS)

    The AARE [Advanced Analysis for Reactor Engineering] modular code system for physics analysis of a wide range of reactor systems, including fusion blankets and shielding, is being developed jointly between PSI [Paul Scherrer Institute] and the Los Alamos National Laboratory. In the present work, a general description of the system, its philosophy and capabilities, together with a brief summary of its modules and data libraries, is given. The cross-section shielding and reformatting code, TRAMIX, is described in more detail

  14. Advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    This paper re-examines the rationale for advanced nuclear fuel cycles in general, and for CANDU advanced fuel cycles in particular. The traditional resource-related arguments for more uranium nuclear fuel cycles are currently clouded by record-low prices for uranium. However, the total known conventional uranium resources can support projected uranium requirements for only another 50 years or so, less if a major revival of the nuclear option occurs as part of the solution to the world's environmental problems. While the extent of the uranium resource in the earth's crust and oceans is very large, uncertainty in the availability and price of uranium is the prime resource-related motivation for advanced fuel cycles. There are other important reasons for pursuing advanced fuel cycles. The three R's of the environmental movement, reduce, recycle, reuse, can be achieved in nuclear energy production through the employment of advanced fuel cycles. The adoption of more uranium-conserving fuel cycles would reduce the amount of uranium which needs to be mined, and the environmental impact of that mining. Environmental concerns over the back end of the fuel cycle can be mitigated as well. Higher fuel burnup reduces the volume of spent fuels which needs to be disposed of. The transmutation of actinides and long-lived fission products into short-lived fission products would reduce the radiological hazard of the waste from thousands to hundreds of years. Recycling of uranium and/or plutonium in spent fuel reuses valuable fissile material, leaving only true waste to be disposed of. Advanced fuel cycles have an economical benefit as well, enabling a ceiling to be put on fuel cycle costs, which are

  15. Advanced fuel cycles of WWER-1000 reactors

    International Nuclear Information System (INIS)

    The present paper considers characteristics of fuel cycles for the WWER-1000 reactor satisfying the following conditions: duration of the campaign at the nominal power is extended from 250 EFPD up to 470 and more ones; fuel enrichment does not exceed 5 wt.%; fuel assemblies maximum burnup does not exceed 55 MWd/kgHM. Along with uranium fuel, the use of mixed Uranium-Plutonium fuel is considered. Calculations were conducted by codes TVS-M, BIPR-7A and PERMAK-A developed in the RRC Kurchatov Institute, verified for the calculations of uranium fuel and certified by GAN RF

  16. 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)

  17. Thorium utilization in ACR (Advanced CANDU) and CANDU-6 reactors

    International Nuclear Information System (INIS)

    It is the main objective of this study to investigate fuel composition options for CANDU type of reactors that are capable of using a mixture of U-Th as fuel. A homogenous mixture of (U-Th)O2 was used in all elements of fuel bundles. The core of CANDU-6 and ACR (Advanced CANDU) were modeled using MCNP5. In equilibrium core, using MONTEBURNS2 code (coupled with MCNP5 and ORIGENS) for once-through uranium and once-through uranium-thorium fuel cycle of CANDU-6 and ACR, discharge burnups and spent fuel compositions were computed. For various enrichments of uranium and different fractions of thorium in a uranium-thorium fuel mixture, performing burnup calculations, relevant relations were derived; in addition, conversion ratio, fuel requirement, uranium resource utilization, and natural uranium savings were determined, and their changes with burnup were observed. Appropriate fuel compositions were discussed.

  18. Assessment of United States industry structural codes and standards for application to advanced nuclear power reactors: Appendices. Volume 2

    International Nuclear Information System (INIS)

    Throughout its history, the USNRC has remained committed to the use of industry consensus standards for the design, construction, and licensing of commercial nuclear power facilities. The existing industry standards are based on the current class of light water reactors and as such may not adequately address design and construction features of the next generation of Advanced Light Water Reactors and other types of Advanced Reactors. As part of their on-going commitment to industry standards, the USNRC commissioned this study to evaluate US industry structural standards for application to Advanced Light Water Reactors and Advanced Reactors. The initial review effort included (1) the review and study of the relevant reactor design basis documentation for eight Advanced Light Water Reactors and Advanced Reactor Designs, (2) the review of the USNRCs design requirements for advanced reactors, (3) the review of the latest revisions of the relevant industry consensus structural standards, and (4) the identification of the need for changes to these standards. The results of these studies were used to develop recommended changes to industry consensus structural standards which will be used in the construction of Advanced Light Water Reactors and Advanced Reactors. Over seventy sets of proposed standard changes were recommended and the need for the development of four new structural standards was identified. In addition to the recommended standard changes, several other sets of information and data were extracted for use by USNRC in other on-going programs. This information included (1) detailed observations on the response of structures and distribution system supports to the recent Northridge, California (1994) and Kobe, Japan (1995) earthquakes, (2) comparison of versions of certain standards cited in the standard review plan to the most current versions, and (3) comparison of the seismic and wind design basis for all the subject reactor designs

  19. Assessment of United States industry structural codes and standards for application to advanced nuclear power reactors: Final report. Volume 1

    International Nuclear Information System (INIS)

    Throughout its history, the USNRC has remained committed to the use of industry consensus standards for the design, construction, and licensing of commercial nuclear power facilities. The existing industry standards are based on the current class of light water reactors and as such may not adequately address design and construction features of the next generation of Advanced Light Water Reactors and other types of Advanced Reactors. As part of their on-going commitment to industry standards, the USNRC commissioned this study to evaluate US industry structural standards for application to Advanced Light Water Reactors and Advanced Reactors. The initial review effort included: (1) the review and study of the relevant reactor design basis documentation for eight Advanced Light Water Reactors and Advanced Reactor Designs, (2) the review of the USNRCs design requirements for advanced reactors, (3) the review of the latest revisions of the relevant industry consensus structural standards, and (4) the identification of the need for changes to these standards. The results of these studies were used to develop recommended changes to industry consensus structural standards which will be used in the construction of Advanced Light Water Reactors and Advanced Reactors. Over seventy sets of proposed standard changes were recommended and the need for the development of four new structural standards was identified. In addition to the recommended standard changes, several other sets of information and data were extracted for use by USNRC in other ongoing programs. This information included: (1) detailed observations on the response of structures and distribution system supports to the recent Northridge, California (1994) and Kobe, Japan (1995) earthquakes, (2) comparison of versions of certain standards cited in the standard review plan to the most current versions, and (3) comparison of the seismic and wind design basis for all the subject reactor designs

  20. 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.)

  1. Foundational development of an advanced nuclear reactor integrated safety code.

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  2. Qualification issues for advanced light-water reactor protection systems

    International Nuclear Information System (INIS)

    The instrumentation and control (I ampersand C) systems in advanced reactors will make extensive use of digital controls, microprocessors, multiplexing, and fiber optic transmission. Elements of these advances in I ampersand C have been implemented on some current operating plants. However, the widespread use of the above technologies, as well as the use of artificial intelligence with minimum reliance on human operator control of reactors, highlights the need to develop standards for qualifying the I ampersand C used in the next generation of nuclear power plants. As a first step in this direction, the protection system I ampersand C for present-day plants was compared to that proposed for advanced light-water reactors (ALWRs). An evaluation template was developed by assembling a configuration of a safety channel instrument string for a generic ALWR, then comparing the impact of environmental stressors on that string to their effect on an equivalent instrument string from an existing light-water reactor. The template was then used to suggest a methodology for the qualification of microprocessor-based protection systems. The methodology identifies standards/regulatory guides (or lack thereof) for the qualification of microprocessor-based safety I ampersand C systems. This approach addresses in part issues raised in NRC policy document SECY-91-292, which recognizes that advanced I ampersand C systems for the nuclear industry are ''being developed without consensus standards. as the technology available for design is ahead of the technology that is well understood through experience and supported by application standards.''

  3. 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.

  4. Advanced gas cooled nuclear reactor materials evaluation and development program

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.

  5. Advanced gas cooled nuclear reactor materials evaluation and development program

    International Nuclear Information System (INIS)

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed

  6. Advanced monitoring and control systems for fast reactors

    International Nuclear Information System (INIS)

    One of the important aspects of nuclear power station (NPS) improvement with fast reactors is provision of safety. The safety conception of advanced fast power reactors is directed on elaborating such solutions where as much as possible properties of reactor self-protection and natural laws are used in which the self-protection of the nuclear reactor is realized. To these solutions we may refer the usage of hydraulically weighted rods of alarm protection, negative temperature and power coefficients, negative sodium empty effect, natural circulation without power sources, natural convection and other measures. Additionally special technological systems are envisaged, which start functioning with the coming of the initial event of the accident. 1 ref., 7 figs, 1 tab

  7. 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

  8. 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

  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. (author)

  10. 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

  11. Liquid sloshing in gravity driven water pool of Advanced Heavy Water Reactor - pool liquid under design seismic load and slosh control studies

    International Nuclear Information System (INIS)

    Sloshing phenomenon is well understood in regular cylindrical and rectangular liquid tanks subjected to earthquake. However, seismic behaviour of water in complex geometry such as a sectored annular tank, e.g., Gravity Driven Water Pool (GDWP) which is located in Advanced Heavy Water Reactor (AHWR) need to be investigated in detail in the view of safety significance. Initially, for validation of Computational Fluid Dynamics (CFD) procedure, square and four sectored square tanks are taken. Slosh height and liquid pressure are calculated over time through theoretical and experimental procedures. Results from theoretical and experimental approaches are compared with CFD results and found to be in agreement. The present work has two main objectives. The first one is to investigate the sloshing behaviour in an un-baffled and baffled three dimensional single sector of GDWP of AHWR under sinusoidal excitation. Other one is to study the sloshing in GDWP water using simulated seismic load along the three orthogonal directions. This simulated seismic load is generated from design basis floor response spectrum data (FRS) of AHWR building. For this, the annular tank is modelled along with water and numerical simulation is carried out. The sinusoidal and earthquake excitations are applied as acceleration force along with gravity. For the earthquake case, acceleration-time history is generated compatible to the design FRS of AHWR building. The free surface is captured by Volume of Fluid (VOF) technique and the fluid domain is solved by finite volume method while the structural domain is solved by finite element approach. Un-baffled and baffled tank configurations are compared to show the reduction in wave height under excitation. The interaction between the fluid and pool wall deformation is simulated using a partitioned fluid-structure coupling. In the earthquake case, a user subroutine function is developed to convert FRS in to time history of acceleration in three directions

  12. Alternate-fuel reactor studies

    International Nuclear Information System (INIS)

    A number of studies related to improvements and/or greater understanding of alternate-fueled reactors is presented. These studies cover the areas of non-Maxwellian distributions, materials and lifetime analysis, a 3He-breeding blanket, tritium-rich startup effects, high field magnet support, and reactor operation spanning the range from full D-T operation to operation with no tritium breeding

  13. INITIAL IRRADIATION OF THE FIRST ADVANCED GAS REACTOR FUEL DEVELOPMENT AND QUALIFICATION EXPERIMENT IN THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti

    2007-09-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors 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 United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. 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, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  14. Commercial scale performance predictions for high-temperature electrolysis plants coupled to three advanced reactor types

    International Nuclear Information System (INIS)

    This paper presents results of system analyses that have been developed to assess the hydrogen-production performance of commercial-scale high-temperature electrolysis (HTE) plants driven by three different advanced reactor - power-cycle combinations: a high-temperature helium-cooled reactor coupled to a direct Brayton power cycle, a supercritical CO2-cooled reactor coupled to a direct recompression cycle, and a sodium-cooled fast reactor coupled to a Rankine cycle. The system analyses were performed using UniSim software. The work described in this report represents a refinement of previous analyses in that the process flow diagrams include realistic representations of the three advanced reactors directly coupled to the power cycles and integrated with the high-temperature electrolysis process loops. In addition, this report includes parametric studies in which the performance of each HTE concept is determined over a wide range of operating conditions. Results of the study indicate that overall thermal-to-hydrogen production efficiencies (based on the low heating value of the produced hydrogen) in the 45 - 50% range can be achieved at reasonable hydrogen production rates with the high-temperature helium-cooled reactor concept, 42 - 44% with the supercritical CO2-cooled reactor and about 33 - 34% with the sodium-cooled reactor. (authors)

  15. Metal fire implications for advanced reactors. Part 1, literature review

    International Nuclear Information System (INIS)

    Public safety and acceptance is extremely important for the nuclear power renaissance to get started. The Advanced Burner Reactor and other potential designs utilize liquid sodium as a primary coolant which provides distinct challenges to the nuclear power industry. Fire is a dominant contributor to total nuclear plant risk events for current generation nuclear power plants. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation. An advanced understanding of metal fire behavior in regards to the new designs will benefit both science and industry. This report presents an extensive literature review that captures past experiences, new advanced reactor designs, and the current state-of-knowledge related to liquid sodium combustion behavior

  16. Metal fire implications for advanced reactors. Part 1, literature review.

    Energy Technology Data Exchange (ETDEWEB)

    Nowlen, Steven Patrick; Radel, Ross F.; Hewson, John C.; Olivier, Tara Jean; Blanchat, Thomas K.

    2007-10-01

    Public safety and acceptance is extremely important for the nuclear power renaissance to get started. The Advanced Burner Reactor and other potential designs utilize liquid sodium as a primary coolant which provides distinct challenges to the nuclear power industry. Fire is a dominant contributor to total nuclear plant risk events for current generation nuclear power plants. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation. An advanced understanding of metal fire behavior in regards to the new designs will benefit both science and industry. This report presents an extensive literature review that captures past experiences, new advanced reactor designs, and the current state-of-knowledge related to liquid sodium combustion behavior.

  17. Identification of improvements of advanced light water reactor concepts

    International Nuclear Information System (INIS)

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

  18. The core design of the advanced power reactor plus (APR+)

    International Nuclear Information System (INIS)

    Advance Power Reactor Plus (APR+), a pressurized water reactor and an improved nuclear power reactor based on the Advanced Power Reactor 1400 MWe (APR1400) in Korea, has been developed with 18-month cycle operation strategy from its initial core. The APR+ core power is 4290 MWth which corresponds to a 1500 MWe class nuclear power plant. The reactor core consists of 257 fuel assemblies. Comparing with APR1400 core design, 16 fuel assemblies are added. Its cycle length is expected about 450 EFPD directly from initial core, although most of previous other plants had been started according to their annual or 15-month cycle operation schedule at their initial core and gone to 18-month after third - fourth cycle. In order to reduce the peaking power, fuel pin configurations of the assembly, are optimized by using some low enriched fuel pins and gadolinia bearings. APR+ core has been met the requirements as well as the above cycle length requirement; 1) peaking factor, 2) Negative MTC(Moderator Temperature Coefficient), 3) sufficient shutdown margin, 4) convergent Xenon stability Index. The maximum rod burnup and the discharge fuel assembly burnup are also satisfied those of the limit. It is expected to acquire the standard design approval by the end of 2012 by the Korean nuclear regulatory. (authors)

  19. 20% inlet header break analysis of Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    The proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. A passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power levels, with no primary coolant pumps. Loss of coolant due to failure of inlet header results in depressurization of primary heat transport (PHT) system and containment pressure rise. Depressurization activates various protective and engineered safety systems like reactor trip, isolation condenser and advanced accumulator, limiting the consequences of the event. This paper discusses the thermal hydraulic transient analysis for evaluating the safety of the reactor, following 20% inlet header break using RELAP5/MOD3.2. For the analysis, the system is discretized appropriately to simulate possible flow reversal in one of the core paths during the transient. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, steam quality and flow in different parts of the Primary Heat Transport (PHT) system. Flow and energy discharges into the containment are also estimated for use in containment analysis. (author)

  20. Helias reactor studies

    International Nuclear Information System (INIS)

    The Helias reactor is an upgraded version of the Wendelstein 7-X experiment. The magnetic field has 5 field periods and the main optimization principle is the reduction of the Pfirsch-Schlueter currents and the Shafranov shift, which has been verified by computations with the NEMEC and MFBE-codes. The modular coil system comprises 50 coils, which are constructed using NbTi-superconducting cables. The basic dimensions are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 5 T, maximum field on the coils 10 T. Forces and stresses in the coil system have been investigated with the aid of the ANSYS code, which found maximum stress values of about 650 MPa in the coil casing. Helias configurations with 4 and 3 field periods have been constructed by starting from the 5-period case and by eliminating one or two periods while the shape of the coils is kept nearly invariant. In a first survey blanket concepts, developed for the DEMO tokamak, have been adapted to the Helias geometry, in particular, the solid breeder concept developed by FZK (Karlsruhe) has been extrapolated to the Helias geometry identifying the drawbacks and advantages of this concept. Furthermore, the liquid breeder concept using Li7-Pb83 and water-cooling is an interesting alternative for the Helias reactor. Maintenance of blanket and plasma facing components is possible through the portholes between modular coils. Numerical simulations of the start-up phase of the Helias reactor using the TOTAL-P code have confirmed the zero-dimensional modeling of the fusion plasma with the aid of empirical scaling laws. (author)

  1. Proceedings of DAE-BRNS theme meeting on advances in reactor physics: design, analysis and operation of nuclear reactors

    International Nuclear Information System (INIS)

    Reactor Physics forms a very vital part of Nuclear Engineering. It often initiates and contributes to new reactor concepts and designs. Reactor physics starts with nuclear physics aspects in the form of nuclear data required for analysis and interacts intensively with engineering aspects of design. Apart from design, physics aspects of operation, safety analysis, fuel management, fuel cycle studies etc are areas in which contributions are demanded from reactor physics. Efficient, economic and safe operation of any nuclear reactor requires good understanding of the physics aspects, in terms of operating regimes, available control and safety margins all the time. Due to such a central role, regulatory aspects of reactor physics are also important. For these tasks efficient and accurate modeling and computation tools are vital. The theoretical computations require experimental validation, which again requires special capabilities in nuclear detectors, electronics and analysis. The discussions in this theme meeting are on the recent advancements in reactor physics with reference to design, safety and operation. Papers relevant to INIS are indexed separately

  2. 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.

  3. The Consortium for Advanced Simulation of Light Water Reactors

    International Nuclear Information System (INIS)

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

  4. The Consortium for Advanced Simulation of Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ronaldo Szilard; Hongbin Zhang; Doug Kothe; Paul Turinsky

    2011-10-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

  5. Characterization of plasma-induced phenol advanced oxidation process in a DBD reactor

    OpenAIRE

    Marotta, E.; Ceriani, E.; Shapoval, V.; Schiorlin, M.; Ceretta, C.; Rea, M.; Paradisi, C.

    2011-01-01

    Abstract Using phenol as a model organic pollutant we studied and characterized an innovative advanced oxidation process in water using a prototype dielectric barrier discharge (DBD) reactor in which electrical discharges are produced in the air above the water surface. Phenol is decomposed quite efficiently in this reactor operated at room temperature and atmospheric pressure. The process selectivity to form CO2 is, however, to be improved since a large fraction of the treated org...

  6. 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

  7. Reflector modeling in advanced nodal analysis of pressurized water reactors

    International Nuclear Information System (INIS)

    Recent progress in the modeling of the reflector regions of pressurized water reactors within the framework of advanced nodal diffusion analysis methods is reviewed. Attention is focused on the modeling of the radial reflector of a PWR which is most problematic because of its irregular and heterogeneous structure. Numerical results are presented to demonstrate the high accuracy of the methods which are now available for generating nodal reflector parameters and it is shown that errors due to reflector modeling in multi-dimensional nodal reactor analysis can be practically eliminated. (author). 23 refs, 1 fig., 2 tabs

  8. Economics of advanced light water reactors - Recent update

    International Nuclear Information System (INIS)

    This paper includes recently updated analyses of the economic prospects of advanced light water reactors (ALWRs) during the decade of the 1990s. United Engineers and Constructors (UE and C) has performed engineering economic analyses related to ALWRs over the last 5 yr using both target economics and detailed cost-estimating methodologies. It has been found through such cost comparisons that properly designed and constructed ALWRs should cost less than the target cost figures listed above and significantly less than the pressurized water reactor better experience reference LWR plant cost

  9. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-04-15

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

  10. Description of the advanced gas cooled type of reactor (AGR)

    International Nuclear Information System (INIS)

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs

  11. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-01-20

    ... COMMISSION 10 CFR Part 52 RIN 3150-AI84 U.S. Advanced Boiling Water Reactor Aircraft Impact Design... the U.S. Advanced Boiling Water Reactor (ABWR) standard plant design to comply with the NRC's aircraft...--Design Certification Rule for the U.S. Advanced Boiling Water Reactor IV. Section-by-Section Analysis...

  12. Station blackout core damage frequency in an advanced nuclear reactor

    International Nuclear Information System (INIS)

    Even though nuclear reactors are provided with protection systems so that they can be automatically shut down in the event of a station blackout, the consequences of this event can be severe. This is because many safety systems that are needed for removing residual heat from the core and for maintaining containment integrity, in the majority of the nuclear power plants, are AC dependent. In order to minimize core damage frequency, advanced reactor concepts are being developed with safety systems that use natural forces. This work shows an improvement in the safety of a small nuclear power reactor provided by a passive core residual heat removal system. Station blackout core melt frequencies, with and without this system, are both calculated. The results are also compared with available data in the literature. (author)

  13. Advanced heavy water reactor pressure tube-easy replaceability

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor (AHWR) is a 300 MWe vertical pressure tube type reactor. A coolant channel consists of pressure tube, made of Zr-2.5 % Nb, which is separated from cold calandria tube using garter spring spacers. The principal function of pressure tube is to support and locate the fuel assembly and allows light water coolant through fuel assembly by natural circulation. Since AHWR is designed for life of 100 years, it necessitates the replacement of pressure tubes during service life. Easy replaceability of pressure tube, along with surveillance requirements, has major bearing on the design of coolant channel assembly. The several systems and tools have been conceptualised to cater the needs for easy and quick replacement of a pressure tube during reactor shut down. This paper gives the highlights of the innovative design features of coolant channel, preliminary design and pre-requisites for replacement, and experimental programme for demonstration of easy replaceability. (author)

  14. Incorporating outage management principles into the advanced light water reactor

    International Nuclear Information System (INIS)

    In the United States there are 110 light water reactor (LWR) plants currently in operation, with a total generating capacity of 102 580 MW(electric). These plants include 37 boiling water reactor (BWR) and 73 pressurized water reactor (PWR) units. Since 1980, more than 40 nuclear power plants have entered service in the United States. However, no new plants have been ordered by utilities and owners groups since 1978. There will come a time in the not-too-distant future that new, large electricity generating units will be needed to supply expected increases in base-load capacity. Will the new advanced LWR (ALWR) designs be able to pass muster and be chosen to help meet that need? With outage management at operating plants improving every year, what can the ALWR designs offer that has not already been incorporated?

  15. Water chemistry features of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor (AHWR) being designed in India proposes to use Plutonium and Thorium as fuel. The objective is to extract energy from the uranium-233 formed from Thorium. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a natural circulation reactor. Thus, it has got several advanced passive safety features built into the system. The various water coolant systems are listed below. i) Main Heat transport System ii) Feed water system iii) Condenser cooling system iv) Process water system and safety systems. As it is a tube type reactor, the radiolysis control differs from the normal boiling water reactor. The coolant enters the bottom of the coolant channel, boiling takes place and then the entire steam water mixture exits the core through the long tail pipes and reaches the moisture separator. Thus, there is a need to devise methods to protect the tail pipes from oxidizing water chemistry condition. Similarly, the moderator heavy water coolant chemistry differs from that of moderator system chemistry of PHWR. The reactivity worth per ppm of gadolinium and boron are low in comparison to PHWR. As a result, much higher concentration of neutron poison has to be added for planned shutdown, start up and for actuating SDS-2. The addition of higher concentration of neutron poison result in higher radiolytic production of deuterium and oxygen. Their recombination back to heavy water has to take into account the higher production of these gases. This paper also discusses the chemistry features of safety systems of AHWR. In addition, the presentation will cover the chemistry monitoring methodology to be implemented in AHWR. (author)

  16. Advanced-fuel reversed-field pinch reactor (RFPR)

    International Nuclear Information System (INIS)

    The utilization of deuterium-based fuels offers the potential advantages of greater flexibility in blanket design, significantly reduced tritium inventory, potential reduction in radioactivity level, and utilization of an inexhaustible fuel supply. The conventional DT-fueled Reversed-Field Pinch Reactor (RFPR) designs are reviewed, and the recent extension of these devices to advanced-fuel (catalyzed-DD) operation is presented. Attractive and economically competitive DD/RFPR systems are identified having power densities and plasma parameters comparable to the DT systems. Converting an RFP reactor from DT to DD primarily requires increasing the magnetic field levels a factor of two, still requiring only modest magnet coil fields (less than or equal to 4 T). When compared to the mainline tokamak, the unique advantages of the RFP (e.g., high beta, low fields at the coils, high ohmic-heating power densities, unrestricted aspect ratio) are particularly apparent for the utilization of advanced fuels

  17. Toxicity of irradiated advanced heavy water reactor fuels.

    Science.gov (United States)

    Priest, N D; Richardson, R B; Edwards, G W R

    2013-02-01

    The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels. PMID:23274823

  18. ORNL R and D on advanced small and medium power reactors: Selected topics

    International Nuclear Information System (INIS)

    From 1984-1985, ORNL studied several innovative small and medium power nuclear concepts with respect to viability. Criteria for assessment of market attractiveness were developed and are described here. Using these criteria and descriptions of selected advanced reactor concepts, and assessment of their projected market viability in the time period 2000-2010 was made. All of these selected concepts could be considered as having the potential for meeting the criteria but, in most cases, considerable RandD would be required to reduce uncertainties. This work and later studies of safety and licensing of advanced, passively safe reactor concepts by ORNL are described. The results of these studies are taken into account in most of the current (FY 1989) work at ORNL on advanced reactors. A brief outline of this current work is given. One of the current RandD efforts at ORNL which addresses the operability and safety of advanced reactors is the Advanced Controls Program. Selected topics from this Program are described. 13 refs., 1 fig

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

    International Nuclear Information System (INIS)

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

  20. The DOE Advanced Gas Reactor Fuel Development and Qualification Program

    International Nuclear Information System (INIS)

    The high outlet temperatures and high thermal-energy conversion efficiency of modular High Temperature Gas-cooled Reactors (HTGRs) enable an efficient and cost effective integration of the reactor system with non-electricity generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300 C and 900 C. The Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission-product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete, fundamental understanding of the relationship between the fuel fabrication process and key fuel properties, the irradiation and accident safety performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. An overview of the program and recent progress is presented.

  1. Johnson Noise Thermometry for Advanced Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Britton Jr, Charles L [ORNL; Roberts, Michael [ORNL; Bull, Nora D [ORNL; Holcomb, David Eugene [ORNL; Wood, Richard Thomas [ORNL

    2012-10-01

    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 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 this project is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

  2. 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

  3. Formulation of a possible advanced reactor legislative strategy and proposal

    International Nuclear Information System (INIS)

    A number of initiatives have been taken to date regarding the formulation of legislation to support in various ways the DOE advanced nuclear reactor program. Among the more prominent of these are bills that have been introduced by Sen. Johnston (D-La) and Rep. Udall (D-Az) as well as a draft bill put together by the nuclear industry and that could be introduced by Rep. Stallings (D-Id). These legislative initiatives are presented in this paper

  4. Advanced technologies for water cooled reactors 1990. Pt. 1

    International Nuclear Information System (INIS)

    The meeting was attended by 20 participants from 12 countries who reviewed and discussed the status and progress of national programmes on advanced water-cooled reactors and recommended to the Scientific Secretary a comprehensive programme for 1991/1992 which would support technology development programmes in IWGATWR Member States. This summary report outlines the activities of IWGATWR since its Second Meeting in June 1988 and main results of the Third Meeting

  5. Advanced technologies for water cooled reactors 1990. Pt. 2

    International Nuclear Information System (INIS)

    The main purpose of the meeting was to review and discuss the status of national programmes, the progress achieved since the last meeting held in June 1988 in the field of advanced technologies and design trends for existing and future water cooled reactors. 24 specialists from 14 countries and the IAEA took part in the meeting and 12 papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  6. Advancing the CANDU reactor: From generation to generation

    International Nuclear Information System (INIS)

    Emphasizing safety, reliability and economics, the CANDU reactor development strategy is one of continuous improvement, offering value and assured support to customers worldwide. The Advanced CANDU Reactor (ACR-1000) generation, designed by Atomic Energy of Canada Limited (AECL), meets the new economic expectation for low-cost power generation with high capacity factors. The ACR is designed to meet customer needs for reduced capital cost, shorter construction schedule, high plant capacity factor, low operating cost, increased operating life, simple component replacement, enhanced safety features, and low environmental impact. The ACR-1000 design evolved from the internationally successful medium-sized pressure tube reactor (PTR) CANDU 6 and incorporates operational feedback from eight utilities that operate 31 CANDU units. This technical paper provides a brief description of the main features of the ACR-1000, and its major role in the development path of the generations of the pressure tube reactor concept. The motivation, philosophy and design approach being taken for future generation of CANDU pressure tube reactors are described

  7. 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.

  8. 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.

  9. Nordic study on reactor waste

    International Nuclear Information System (INIS)

    In 1981, 14 nuclear power reactors are in operation and 2 under construction in the Nordic countries. So far, the reactor waste originating from day-to-day operation of these plants has been stored in solidified form at the reactor sites. Within a few years a satisfactory disposal procedure needs to be established. While the main R and D effects in the waste field have earlier been devoted to the question of irradiated fuel and waste from reprocessing, there is therefore now an increased interest in reactor waste with its much lower radioactivity but somewhat larger volumes. Since 1977, efforts have been made in a joint Nordic study to examine which facts need to be known in order to perform a comprehensive safety assessment of a reactor waste management system. In the present study a Reference system related to the waste generated over 30 years from six 500 MW-reactors is examined. The dominating radionuclides during storage and transportation accident scenarios are Cs-134, Cs-137 and Co-60. For most of the release scenarios from repositories Cs-137 and Sr-90 are dominating. Some scenarios are, however, dominated by the very longlived nuclides I-129 and C-14. A closer examination of the concentration in the waste of these nuclides and of their leaching properties indicates that their small - but significant - influence, as calculated, is probably grossly overestimated. The mechanical stability obtained in routine solidification processes of reactor waste products in conjunction with the outer container (steel drum, transport container, etc.) turns out to be sufficient. Difficulties were encountered in applying ICRP methodology and available dose calculation methods to calculation of population doses due to small activity releases, and effects extending into the far future. (EG)

  10. Tokamak experimental power reactor studies

    International Nuclear Information System (INIS)

    The principal results of a scoping and project definition study for the Tokamak Experimental Power Reactor are presented. Objectives are discussed; a preliminary conceptual design is described; detailed parametric, survey and sensitivity studies are presented; and research and development requirements are outlined. (U.S.)

  11. 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

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

  12. Advanced Test Reactor National Scientific User Facility Partnerships

    International Nuclear Information System (INIS)

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

  13. US reactor safety study reviewed

    International Nuclear Information System (INIS)

    The policy statement issued by the United States Nuclear Regulatory Commission concerning its views of the Reactor Safety Study (WASH-1400) in the light of criticisms of the study made by a special review group headed by Dr Harold Lewis, is discussed. Dr Lewis appeared before a Congressional sub-committee to discuss the findings of the review group and his testimony is reproduced here. This is considered under the four headings; safety - what is it, WASH-1400 and RARG - what was found, risk assessment methodology -where do we go from here, and are reactors safe. (UK)

  14. Advanced epithermal thorium reactor (AETR) physics

    International Nuclear Information System (INIS)

    The AETR concept is reviewed in reference to existing theory, nuclear parameters, and potential neutron economy. The effect of thorium resonance capture in graphite-moderate d systems with median absorption energies in the range from 0.10 to 100 keV have been studied. Narrow-resonance (NR) and wide-resonance (NRIA) formulations are used to obtain the temperature-dependent effective resonance integral of thorium rods which are expressed as equivalent multi-group cross-sections. The need for nuclear data in the intermediate energy range led to design and con- struction of a critical assembly. Nuclear design of this assembly emphasizes the importance of cross-section data and the theoretical interpretation of these experimental results, both pertinent to the design of an AETR. The accuracy of the analytical techniques has been demonstrated in the analysis of ZPR-III experimental results. Three heat-transfer configurations are compared using doubling time as an optimizing parameter. The effect of Pa233 and uranium-isotope s production on relative neutron economy, potential breeding ratios, and burn-up characteristics are evaluated in regard to the uncertainties in the nuclear cross-sections. (author)

  15. 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

  16. Achievements and prospects for advanced reactor design and fuel cycles

    International Nuclear Information System (INIS)

    The future of Nuclear Energy relies on the complementary optimization of reactors for NPPs and the associated nuclear fuel cycles. This is an apparent contradiction if we look in the so large effort made worldwide for developing advance reactors for power plants alone. The vision that focus the optimization effort in reactors and in the other side and separated in the associated fuel cycle jeopardizes the final results of an optimized nuclear system. The control of the primary source of energy is a key question and the technology involved and its control the main issue to be considered when the evaluation of advanced nuclear systems are under consideration. However the main reason of this situation is that reactors for NPP is still been costly, inefficient compared with other energy converters and increasingly complex to accomplish safety requirements. The maturity of nuclear technology and the present NPP are the background for the evolutionary concepts of reactors while the response to economy, safety, waste generation and management and proliferation resistance are the drivers for innovative concepts. Most traditional technology holders and NPP vendors have evolutionary LWR and HWR systems and participate directly or indirectly in innovative projects for future applications including fast reactors. EPR, AP 1000, KSNP, ABWR, WWER-600, ACR-700 and AHWR are examples of this fact. Example of continuous effort in fast reactors development are MONJU reactor, CEFR, FBTR and the emblematic Superphenix. Both reactors and nuclear fuel cycles should evolve throughout a breakthrough process if the energy demand mainly becomes large in developing countries. This may require a different approach that the one that drives the past 50 years mainly because the modules should be optimized for quite different electricity markets. Small and Medium Power Reactors like SMART, CAREM, IRIS, PBMR and HTGRs, enrichment processes optimized to be economics for small capacity production

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

    International Nuclear Information System (INIS)

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

  18. Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2006-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors 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. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. 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 goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  19. Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The United States Department of Energy's Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors 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. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. 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 goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation

  20. Qualification of the reactor physics toolset for the design and analysis of the advanced CANDU reactor

    International Nuclear Information System (INIS)

    The qualification of reactor physics toolset for Advanced CANDU Reactor (ACR) applications is described in this paper. The qualification process follows AECL standard code validation methodology. The ACR nuclear design incorporates certain features that challenge the physics code-suite capabilities. The physics codes were first assessed, and development work required to meet these challenges was undertaken. A Validation Matrix Document was prepared to identify the physics phenomena that could arise during postulated accident events, and specify the experimental data required for code validation. Key issues related to physics modelling and code validation are also discussed. (author)

  1. Vertical liquid film impinged on advanced reactor downcomer wall

    International Nuclear Information System (INIS)

    Experiments are conducted to study the vertical film width and downcomer gap effects on impingement upon the vertical flat wall simulating the core barrel in the APR-1400 (Advanced Power Reactor 1400 MWe). This work is being carried out in ALPHA (Accelerated Liquid Phase Hydrodynamics Apparatus) at the Seoul National University. The results are to be integrated into a system code THEOS (Thermomechanics Hydrodynamics Eutectics Online Simulator). At larger distances from stagnation to the injection nozzle, the gap effect falls off. The data suggest that, for impingement nearing the stagnation point, the flow velocities can be computed by applying a turbulent correction factor to the laminar value calculated for velocities having the same pressure distribution as that present in the impingement region. To find the correction factor, one needs to obtain the liquid film thickness and breakup factors related to the turbulent flow. This is in contrast to the laminar flow, in which the liquid film thickness is uniform and the breakup is nonexistent. This work is particularly concerned with the liquid film thickness. Previous ALPHA experiments with varying nozzle diameters advocate that the correction factor is found to be a function of the injection speed rather than the Reynolds number. This paper documents on measurement of the radial liquid film thickness fluctuations and their point wise values at the same vertical height. It additionally reports on results of numerical calculation for the liquid film profiles. (authors)

  2. Advances in crack-arrest technology for reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bass, B.R.; Pugh, C.E.

    1988-01-01

    The Heavy-Section Steel Technology (HSST) Program at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission is continuing to improve the understanding of conditions that govern the initiation, rapid propagation, arrest, and ductile tearing of cracks in reactor pressure vessel (RPV) steels. This paper describes recent advances in a coordinated effort being conducted under the HSST Program by ORNL and several subcontracting groups to develop the crack-arrest data base and the analytical tools required to construct inelastic dynamic fracture models for RPV steels. Large-scale tests are being carried out to generate crack-arrest toughness data at temperatures approaching and above the onset of Charpy upper-shelf behavior. Small- and intermediate-size specimens subjected to static and dynamic loading are being developed and tested to provide additional fracture data for RPV steels. Viscoplastic effects are being included in dynamic fracture models and computer programs and their utility validated through analyses of data from carefully controlled experiments. Recent studies are described that examine convergence problems associated with energy-based fracture parameters in viscoplastic-dynamic fracture applications. Alternative techniques that have potential for achieving convergent solutions for fracture parameters in the context of viscoplastic-dynamic models are discussed. 46 refs., 15 figs., 3 tabs.

  3. 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

  4. Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

    Energy Technology Data Exchange (ETDEWEB)

    Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

    2006-12-11

    This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

  5. 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

  6. 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)

  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 (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.

  8. Development and assessment of advanced reactor core protection system

    International Nuclear Information System (INIS)

    An advanced core protection system for a pressurized water reactor, Reactor Core Protection System (RCOPS), was developed by adopting a high performance hardware platform and optimal system configuration. The functional algorithms of the core protection system were also improved to enhance the plant availability by reducing unnecessary reactor trips and increasing operational margin. The RCOPS consists of four independent safety channels providing a two-out-of-four trip logic. The reliability analysis using the reliability block diagram method showed the unavailability of the RCOPS to be lower than the conventional system. The failure mode and effects analysis demonstrated that the RCOPS does not lose its intended safety functions for most failures. New algorithms for the RCOPS functional design were implemented in order to avoid unnecessary reactor trips by providing auxiliary pre-trip alarms and signal validation logic for the control rod position. The new algorithms in the RCOPS were verified by comparing the RCOPS calculations with reference results. The new thermal margin algorithm for the RCOPS was expected to increase the operational margin to the limit for Departure from Nucleate Boiling Ratio (DNBR) by approximately 1%. (author)

  9. 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

  10. Preliminary safety evaluation of the advanced burner test reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, F. E.; Fanning, T. H.; Cahalan, J. E.; Nuclear Engineering Division

    2006-09-15

    Results of a preliminary safety evaluation of the Advanced Burner Test Reactor (ABTR) pre-conceptual design are reported. The ABTR safety design approach is described. Traditional defense-in-depth design features are supplemented with passive safety performance characteristics that include natural circulation emergency decay heat removal and reactor power reduction by inherent reactivity feedbacks in accidents. ABTR safety performance in design-basis and beyond-design-basis accident sequences is estimated based on analyses. Modeling assumptions and input data for safety analyses are presented. Analysis results for simulation of simultaneous loss of coolant pumping power and normal heat rejection are presented and discussed, both for the case with reactor scram and the case without reactor scram. The analysis results indicate that the ABTR pre-conceptual design is capable of undergoing bounding design-basis and beyond-design-basis accidents without fuel cladding failures. The first line of defense for protection of the public against release of radioactivity in accidents remains intact with significant margin. A comparison and evaluation of general safety design criteria for the ABTR conceptual design phase are presented in an appendix. A second appendix presents SASSYS-1 computer code capabilities and modeling enhancements implemented for ABTR analyses.

  11. 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

  12. Balanced Design of Safety Systems of CAREM Advanced Reactor

    International Nuclear Information System (INIS)

    Nuclear Power Plants must meet the performance that the market and the population demand in order to be part of the electricity supply industry.It is related mainly with the results of reactor's economy and safety.New advances in the methodology developed for reactor economic optimization analyzing its safety at an early engineering stage, aiming at balancing these important features of the design, are presented in this work.In particular, the coupling that appears when dimensioning the Emergency Injection System, the Residual Heat Removal System and the containment height of CAREM reactor is described.The new models appended to the computer code that embodies the methodology to balance de designs are shown.Finally the results obtained with the optimizations when applying it are presented.Furthermore, a criterion to establish the maximal diameter for acceptable breaks in RPV's penetrations arises from this work.The application of the methodology and the computer code developed turns out to prove the advantages they provide to reactor design so that the plants are properly balanced and optimized

  13. Advances in the ACR-1000 reactor regulating system and reactor control

    International Nuclear Information System (INIS)

    Advances in the control of the ACR-1000 reactor are presented. The ACR-1000 Reactor Regulating System's (RRS) capability to maintain reactor power at its set point, counteract zonal power deviations, initiate setback as required, and effectively control operational maneuvers including power load-cycling is demonstrated. Three fast core transients and a long Load Cycling transient are presented. For simulations of the fast transients a dynamic RRS Simulation Package (RRS-SP) was developed, where the core neutron kinetics calculations (*CERBERUS module of RFSP) were coupled to a thermal hydraulic code (CATHENA) at every time step. A quasi-static approach was used to demonstrate the RRS performance in the Load Cycling transient that covers five consecutive daily cycles followed by a 2-day weekend cycle. (author)

  14. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  15. Advanced and sustainable fuel cycles for innovative reactor systems

    International Nuclear Information System (INIS)

    The key objective of nuclear energy systems of the future as defined by the Generation IV road map is to provide a sustainable energy generation for the future. It includes the requirement to minimize the nuclear waste produced and thereby notably reduce the long term stewardship burden in the future. It is therefore evident that the corresponding fuel cycles will play a central role in trying to achieve these goals by creating clean waste streams which contain almost exclusively the fission products. A new concept based on a grouped separation of actinides is widely discussed in this context, but it is of course a real challenge to achieve this type of separation since technologies available today have been developed to separate actinides from each other. In France, the CEA has launched extensive research programs in the ATALANTE facility in Marcoule to develop the advanced fuel cycles for new generation reactor systems. In this so called global actinide management (GAM) concept, the actinides are extracted in a sequence of chemical reactions (grouped actinide extraction (GANEX)) and immediately reintroduced in the fuel fabrication process is to use all actinides in the energy production process. The new group separation processes can be derived as in this case from aqueous techniques but also from so-called pyrochemical partitioning processes. Significant progress was made in recent years for both routes in the frame of the European research projects PARTNEW, PYROREP and EUROPART, mainly devoted to the separation of minor actinides in the frame of partitioning and transmutation (P and T) studies. The fuels used in the new generation reactors will be significantly different from the commercial fuels of today. Because of the fuel type and the very high burn-ups reached, pyrometallurgical reprocessing could be the preferred method. The limited solubility of some of the fuel materials in acidic aqueous solutions, the possibility to have an integrated irradiation and

  16. Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti

    2008-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors 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 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, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The design of the first experiment (designated AGR-1) was completed in 2005, and the fabrication and assembly of the test train as well as the support systems and fission product monitoring system that monitor and control the experiment during irradiation were completed in September 2006. The experiment was inserted in the ATR in December 2006, and is serving as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed and the status of the experiment is provided.

  17. Advances in safety engineering for LWR type reactors (retrofitting, re-engineering, EPR, SWR 1000)

    International Nuclear Information System (INIS)

    The paper summarizes the activities of the Siemens company in the field of advanced LWR type reactor engineering and the company's commitments in international projects for the retrofitting and engineered safety improvements of reactor stations in countries of the former Soviet Union. The advances in reactor engineering and the novel design concepts are explained. (orig./CB)

  18. Replacement of the Advanced Test Reactor control room

    International Nuclear Information System (INIS)

    The control room for the Advanced Test Reactor has been replaced to provide modern equipment utilizing current standards and meeting the current human factors requirements. The control room was designed in the early 1960 era and had not been significantly upgraded since the initial installation. The replacement did not change any of the safety circuits or equipment but did result in replacement of some of the recorders that display information from the safety systems. The replacement was completed in concert with the replacement of the control room simulator which provided important feedback on the design. The design successfully incorporates computer-based systems into the display of the plant variables. This improved design provides the operator with more information in a more usable form than was provided by the original design. The replacement was successfully completed within the scheduled time thereby minimizing the down time for the reactor

  19. 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)

  20. Advanced Space Nuclear Reactors from Fiction to Reality

    Science.gov (United States)

    Popa-Simil, L.

    The advanced nuclear power sources are used in a large variety of science fiction movies and novels, but their practical development is, still, in its early conceptual stages, some of the ideas being confirmed by collateral experiments. The novel reactor concept uses the direct conversion of nuclear energy into electricity, has electronic control of reactivity, being surrounded by a transmutation blanket and very thin shielding being small and light that at its very limit may be suitable to power an autonomously flying car. It also provides an improved fuel cycle producing minimal negative impact to environment. The key elements started to lose the fiction attributes, becoming viable actual concepts and goals for the developments to come, and on the possibility to achieve these objectives started to become more real because the theory shows that using the novel nano-technologies this novel reactor might be achievable in less than a century.

  1. 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.

  2. 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

  3. Design of the reactor vessel inspection robot for the advanced liquid metal reactor

    International Nuclear Information System (INIS)

    A consortium of four universities and Oak Ridge National Laboratory designed a prototype wall-crawling robot to perform weld inspection in an advanced nuclear reactor. 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 mock non-hostile environment and shown to perform as expected, as detailed in this report

  4. Evolutionary/advanced light water reactor data report

    International Nuclear Information System (INIS)

    The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (''burned'') in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ''evolutionary'' or ''advanced'' designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ''evolutionary'' LWR alternative

  5. Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

    International Nuclear Information System (INIS)

    The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe2+) irradiation.

  6. Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tan, Lizhen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yang, Ying [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sridharan, K. [Univ. of Wisconsin, Madison, WI (United States)

    2015-12-01

    The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe2+) irradiation.

  7. Pre-licensing of the Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life, and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross electrical output of 1165 MWe. The ACR-1000 design has evolved from AECL's in-depth knowledge of CANDU systems, components, and materials, as well as the experience and feedback received from owners and 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. It also features major improvements in economics, inherent safety characteristics, and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. The CANDU system is ideally suited to this evolutionary approach since the modular fuel channel reactor design can be modified, through a series of incremental changes in the reactor core design, to increase the power output and improve the overall safety, economics, and performance. The safety enhancements made in ACR-1000 encompass improved safety margins, performance and reliability of safety related systems. In particular, the use of the CANFLEX-ACR fuel bundle, with lower linear rating and higher critical heat flux, provides increased operating and safety margins. Safety features draw from those of the existing CANDU plants (e.g., the two

  8. 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

  9. ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    E. Blanford; E. Keldrauk; M. Laufer; M. Mieler; J. Wei; B. Stojadinovic; P.F. Peterson

    2010-09-20

    Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using

  10. ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

    International Nuclear Information System (INIS)

    Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using

  11. Fusion advanced studies Torus

    International Nuclear Information System (INIS)

    The successful development of ITER and DEMO scenarios requires preparatory activities on devices that are smaller than ITER, sufficiently flexible and capable of investigating the peculiar physics of burning plasma conditions. The aim of the Fusion Advanced Studies Torus (FAST) proposal [2.1] (formerly FT3 [2.2]) is to show that the preparation of ITER scenarios and the development of new expertise for the DEMO design and RD can be effectively implemented on a new facility. FAST will a) operate with deuterium plasmas, thereby avoiding problems associated with tritium, and allow investigation of nonlinear dynamics (which are important for understanding alpha particle behaviour in burning plasmas) by using fast ions accelerated by heating and current drive systems; b) work in a dimensionless parameter range close to that of ITER; c) test technical innovative solutions, such as full-tungsten plasma-facing components and an advanced liquid metal divertor target for the first wall/divertor, directly relevant for ITER and DEMO; d) exploit advanced regimes with a much longer pulse duration than the current diffusion time; e) provide a test bed for ITER and DEMO diagnostics; f) provide an ideal framework for model and numerical code benchmarks, their verification and validation in ITER/ DEMO-relevant plasma conditions

  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 fuels for plutonium management in pressurized water reactors

    International Nuclear Information System (INIS)

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h)-1. More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate

  14. Advanced energy system with nuclear reactors as an energy source

    International Nuclear Information System (INIS)

    recovery system is also applicable to a fast reactor (FR) with a supercritical CO2 gas turbine that achieves higher cycle efficiency than conventional sodium cooled FRs with steam turbines. The FR will eliminate problems of conventional FRs related to safety, plant maintenance, and construction costs. The FR consumes efficiently trans-uranium elements (TRU) produced in light water reactors as fuel and reduce long-lived radioactive wastes or environmental loads of long term geological disposal. An Advanced Energy System (AES) with nuclear reactors as an energy source has been proposed which supply electricity and heat to cities. The AES has three objectives: 1. Save energy resources and reduce green house gas emissions, attaining total energy utilization efficiency higher than 85% through waste heat recovery and utilization. 2. Foster a recycling society that produces methane and methanol for fuel cells from waste products of cities and farms. 3. Consume TRU produced in LWRs as fuel for FRs, and reduce long-lived radioactive wastes or environmental loads of long term geological disposal. References 1. Y. Kato, T. Nitawaki and K. Fujima, 'Zero Waste Heat Release Nuclear Cogeneration System, 'Proc. 2003 Intl. Congress on Advanced Nuclear Power Plants (ICAPP'03), Cordoba, Spain, May 4-7, 2003, Paper 3313. 2. Y. Kato, T. Nitawaki and Y. Muto, 'Medium Temperature Carbon Dioxide Gas Turbine Reactor, 'Nucl. Eng. Design, 230, pp. 195-207 (2004). 3. H. N. Tran and Y. Kato, 'New 237Np Burning Strategy in a Supercritical CO2 Cooled Fast Reactor Core Attaining Zero Burnup Reactivity Loss,' Proc. American Nuclear Society's Topical Meeting on Reactor Physics (PHYSOR 2006), Vancouver, British Columbia, Canada, September 10-14, 2006

  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. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    Science.gov (United States)

    2011-10-03

    ... Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... published in the Federal Register on October 21, 2010, (75 FR 65038-65039). Detailed meeting agendas...

  17. Advances in process intensification through multifunctional reactor engineering.

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, Marcia A.; Miller, James Edward; O' Hern, Timothy John; Gill, Walter; Evans, Lindsey R.

    2011-02-01

    A multifunctional reactor is a chemical engineering device that exploits enhanced heat and mass transfer to promote production of a desired chemical, combining more than one unit operation in a single system. The main component of the reactor system under study here is a vertical column containing packing material through which liquid(s) and gas flow cocurrently downward. Under certain conditions, a range of hydrodynamic regimes can be achieved within the column that can either enhance or inhibit a desired chemical reaction. To study such reactors in a controlled laboratory environment, two experimental facilities were constructed at Sandia National Laboratories. One experiment, referred to as the Two-Phase Experiment, operates with two phases (air and water). The second experiment, referred to as the Three-Phase Experiment, operates with three phases (immiscible organic liquid and aqueous liquid, and nitrogen). This report describes the motivation, design, construction, operational hazards, and operation of the both of these experiments. Data and conclusions are included.

  18. Design features of Advanced Power Reactor (APR) 1400 steam generator

    International Nuclear Information System (INIS)

    Advanced Power Reactor 1400 (APR 1400) which is to achieve the improvement of the safety and economical efficiency has been developed by Korea Hydro and Nuclear Power Co., Ltd. (KHNP) with the support from industries and research institutes. The steam generator for APR 1400 is an evolutionary type from System 80+, which is the recirculating U-tube heat exchanger with integral economizer. Compared to the System 80+ steam generator, it is focused on the improved design features, operating and design conditions of APR 1400 steam generator. Especially, from the operation experience of Korean Standard Nuclear Power Plant (KSNP) steam generator, the lessons-learned measures are incorporated to prevent the tube wear caused by flow-induced vibration (FIV). The concepts for the preventive design features against FIV are categorized to two fields; flow distribution and dynamic response characteristics. From the standpoint of flow distribution characteristics, the egg-crate flow distribution plate (EFDP) is installed to prevent the local excessive flow loaded on the most susceptible tube to wear. The parametric study is performed to select the optimum design with the efficient mitigation of local excessive flow. ATHOS3 Mod-01 is used and partly modified to analyze the flow field of the APR 1400 steam generator. In addition, the upper tube bundle support is designed to eliminate the presence of tube with a low natural frequency. Based on the improved upper tube bundle support, the modal analysis is performed and compared with that of System 80+. Using the results of flow distribution and modal analysis, the two mechanisms of flow-induced vibration are investigated; fluid-elastic instability (FEI) and random turbulence excitation (RTE). (authors)

  19. Artificial Intelligent Control for a Novel Advanced Microwave Biodiesel Reactor

    International Nuclear Information System (INIS)

    Biodiesel, an alternative diesel fuel made from a renewable source, is produced by the transesterification of vegetable oil or fat with methanol or ethanol. In order to control and monitor the progress of this chemical reaction with complex and highly nonlinear dynamics, the controller must be able to overcome the challenges due to the difficulty in obtaining a mathematical model, as there are many uncertain factors and disturbances during the actual operation of biodiesel reactors. Classical controllers show significant difficulties when trying to control the system automatically. In this paper we propose a comparison of artificial intelligent controllers, Fuzzy logic and Adaptive Neuro-Fuzzy Inference System(ANFIS) for real time control of a novel advanced biodiesel microwave reactor for biodiesel production from waste cooking oil. Fuzzy logic can incorporate expert human judgment to define the system variables and their relationships which cannot be defined by mathematical relationships. The Neuro-fuzzy system consists of components of a fuzzy system except that computations at each stage are performed by a layer of hidden neurons and the neural network's learning capability is provided to enhance the system knowledge. The controllers are used to automatically and continuously adjust the applied power supplied to the microwave reactor under different perturbations. A Labview based software tool will be presented that is used for measurement and control of the full system, with real time monitoring.

  20. Artificial Intelligent Control for a Novel Advanced Microwave Biodiesel Reactor

    Science.gov (United States)

    Wali, W. A.; Hassan, K. H.; Cullen, J. D.; Al-Shamma'a, A. I.; Shaw, A.; Wylie, S. R.

    2011-08-01

    Biodiesel, an alternative diesel fuel made from a renewable source, is produced by the transesterification of vegetable oil or fat with methanol or ethanol. In order to control and monitor the progress of this chemical reaction with complex and highly nonlinear dynamics, the controller must be able to overcome the challenges due to the difficulty in obtaining a mathematical model, as there are many uncertain factors and disturbances during the actual operation of biodiesel reactors. Classical controllers show significant difficulties when trying to control the system automatically. In this paper we propose a comparison of artificial intelligent controllers, Fuzzy logic and Adaptive Neuro-Fuzzy Inference System(ANFIS) for real time control of a novel advanced biodiesel microwave reactor for biodiesel production from waste cooking oil. Fuzzy logic can incorporate expert human judgment to define the system variables and their relationships which cannot be defined by mathematical relationships. The Neuro-fuzzy system consists of components of a fuzzy system except that computations at each stage are performed by a layer of hidden neurons and the neural network's learning capability is provided to enhance the system knowledge. The controllers are used to automatically and continuously adjust the applied power supplied to the microwave reactor under different perturbations. A Labview based software tool will be presented that is used for measurement and control of the full system, with real time monitoring.

  1. Temperature controlled material irradiation in the advanced test reactor

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) is located at the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho, USA and is owned and regulated by the U.S. Department of Energy (US DOE). The ATR is operated for the US DOE by Lockheed Martin Idaho Technologies. In recent years, prime irradiation space in the ATR has been made available for use by customers having irradiation service needs in addition to the reactor's principal user, the U.S. Naval Nuclear Propulsion Program. To enhance the reactor's capabilities, the US DOE has initiated the development of an Irradiation Test Vehicle (ITV) capable of providing neutron spectral tailoring and temperature control for up to 28 experiments. The ATR-ITV will have the flexibility to simultaneously support a variety of experiments requiring fast, thermal or mixed spectrum neutron environments. Temperature control is accomplished by varying the thermal conductivity across a gas gap established between the experiment specimen capsule wall and the experiment 'in-pile tube (IPT)' inside diameter. Thermal conductivity is adjusted by alternating the control gas mixture ratio of two gases with different thermal conductivities

  2. Artificial Intelligent Control for a Novel Advanced Microwave Biodiesel Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wali, W A; Hassan, K H; Cullen, J D; Al-Shamma' a, A I; Shaw, A; Wylie, S R, E-mail: w.wali@2009.ljmu.ac.uk [Built Environment and Sustainable Technologies Institute (BEST), School of the Built Environment, Faculty of Technology and Environment Liverpool John Moores University, Byrom Street, Liverpool L3 3AF (United Kingdom)

    2011-08-17

    Biodiesel, an alternative diesel fuel made from a renewable source, is produced by the transesterification of vegetable oil or fat with methanol or ethanol. In order to control and monitor the progress of this chemical reaction with complex and highly nonlinear dynamics, the controller must be able to overcome the challenges due to the difficulty in obtaining a mathematical model, as there are many uncertain factors and disturbances during the actual operation of biodiesel reactors. Classical controllers show significant difficulties when trying to control the system automatically. In this paper we propose a comparison of artificial intelligent controllers, Fuzzy logic and Adaptive Neuro-Fuzzy Inference System(ANFIS) for real time control of a novel advanced biodiesel microwave reactor for biodiesel production from waste cooking oil. Fuzzy logic can incorporate expert human judgment to define the system variables and their relationships which cannot be defined by mathematical relationships. The Neuro-fuzzy system consists of components of a fuzzy system except that computations at each stage are performed by a layer of hidden neurons and the neural network's learning capability is provided to enhance the system knowledge. The controllers are used to automatically and continuously adjust the applied power supplied to the microwave reactor under different perturbations. A Labview based software tool will be presented that is used for measurement and control of the full system, with real time monitoring.

  3. Development of Advanced Monitoring System with Reactor Neutrino Detection Technique for Verification of Reactor Operations

    International Nuclear Information System (INIS)

    Recently, technique of Gadolinium-loaded liquid scintillator (Gd-LS) for reactor neutrino oscillation experiments has attracted attention as a monitor of reactor operation and ''nuclear Gain (GA)'' for IAEA safeguards. When the thermal operation power is known, it is, in principle, possible to non-destructively measure the ratio of Pu/U in reactor fuel under operation from the reactor neutrino flux. An experimental program led by Lawrence Livermore National Laboratory and Sandia National Laboratories in USA has already demonstrated feasibility of the reactor monitoring by neutrinos at San Onofre Nuclear Power Station, and the Pu monitoring by neutrino detection is recognized as a candidate of novel technology to detect undeclared operation of reactor. However, further R and D studies of detector design and materials are still necessary to realize compact and mobile detector for practical use of neutrino detector. Considering the neutrino interaction cross-section and compact detector size, the detector must be set at a short distance (a few tens of meters) from reactor core to accumulate enough statistics for monitoring. In addition, although previous reactor neutrino experiments were performed at underground to reduce cosmic ray muon background, feasibility of the measurement at ground level is required for the monitor considering limited access to the reactor site. Therefore, the detector must be designed to be able to reduce external backgrounds extremely without huge shields at ground level, eg. cosmic ray muons and fast neutrons. We constructed a 0.76 ton Gd-LS detector, and carried out a reactor neutrino measurement at the experimental fast reactor JOYO in 2007. The neutrino detector was set up at 24.3m away from the reactor core at the ground level, and we understood the property of the main background; the cosmic-ray induced fast neutron, well. Based on the experience, we are constructing a new detector for the next experiment. The detector is a Gd

  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. 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

  6. MAAP-impair interface for analysis of iodine behavior in advanced reactor accidents

    International Nuclear Information System (INIS)

    As part of the US Department of Energy (US DOE) Advanced Reactor Severe Accident Program, a study was initiated to provide an ex-vessel iodine analytical capability to estimate source terms for severe accidents in advanced light water reactors. This capability has been developed by creating a software link, MID, between the MAAP and IMPAIR computer codes. The interface allows IMPAIR to access the thermal-hydraulic and fission product results provided by MAAP and use these results to drive the chemical reaction and physical mass transfer models in IMPAIR. The first phase of the development is designed to provide iodine analytical capability up to the point of reactor vessel failure. A follow-on study is planned to address iodine behavior in accident scenarios that go beyond vessel failure. A number of MAAP-IMPAIR demonstration calculations have been performed for the General Electric simplified boiling water reactor and Westinghouse AP600 reactor designs. These calculations demonstrated that the software interface provided the necessary link to create a functional ex-vessel iodine analytic capability. They also clearly indicated that both the chemical and the physical behavior of iodine species in the containment are strongly dependent upon the containment thermal-hydraulic conditions

  7. FFTF and Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    GANTT, D.A.

    2000-10-31

    This Resource Load Schedule (RLS) addresses two missions. The Advanced Reactors Transition (ART) mission, funded by DOE-EM, is to transition assigned, surplus facilities to a safe and compliant, low-cost, stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D&D. Facilities to be transitioned include the 309 Building Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy Legacy facilities. This mission is funded through the Environmental Management (EM) Project Baseline Summary (PBS) RL-TP11, ''Advanced Reactors Transition.'' The second mission, the Fast Flux Test Facility (FFTF) Project, is funded through budget requests submitted to the Office of Nuclear Energy, Science and Technology (DOE-NE). The FFTF Project mission is maintaining the FFTF, the Fuels and Materials Examination Facility (FMEF), and affiliated 400 Area buildings in a safe and compliant standby condition. This mission is to preserve the condition of the plant hardware, software, and personnel in a manner not to preclude a plant restart. This revision of the Resource Loaded Schedule (RLS) is based upon the technical scope in the latest revision of the following project and management plans: Fast Flux Test Facility Standby Plan (Reference 1); Hanford Site Sodium Management Plan (Reference 2); and 309 Building Transition Plan (Reference 4). The technical scope, cost, and schedule baseline is also in agreement with the concurrent revision to the ART Fiscal Year (FY) 2001 Multi-Year Work Plan (MYWP), which is available in an electronic version (only) on the Hanford Local Area Network, within the ''Hanford Data Integrator (HANDI)'' application.

  8. Reliability assurance programme guidebook for advanced light water reactors

    International Nuclear Information System (INIS)

    To facilitate the implementation of reliability assurance programmes (RAP) within future advanced reactor programmes and to ensure that the next generation of commercial nuclear reactors achieves the very high levels of safety, reliability and economy which are expected of them, in 1996, the International Atomic Energy Agency (IAEA) established a task to develop a guidebook for reliability assurance programmes. The draft RAP guidebook was prepared by an expert consultant and was reviewed/modified at an Advisory Group meeting (7-10 April 1997) and at a consults meeting (7-10 October 1997). The programme for the RAP guidebook was reported to and guided by the Technical Working Group on Advanced Technologies for Light Water Reactors (TWG-LWR). This guidebook will demonstrate how the designers and operators of future commercial nuclear plants can exploit the risk, reliability and availability engineering methods and techniques developed over the past two decades to augment existing design and operational nuclear plant decision-making capabilities. This guidebook is intended to provide the necessary understanding, insights and examples of RAP management systems and processes from which a future user can derive his own plant specific reliability assurance programmes. The RAP guidebook is intended to augment, not replace, specific reliability assurance requirements defined by the utility requirements documents and by individual nuclear steam supply system (NSSS) designers. This guidebook draws from utility experience gained during implementation of reliability and availability improvement and risk based management programmes to provide both written and diagrammatic 'how to' guidance which can be followed to assure conformance with the specific requirements outlined by utility requirements documents and in the development of a practical and effective plant specific RAP in any IAEA Member State

  9. Preapplication safety evaluation report for the Sodium Advanced Fast Reactor (SAFR) liquid-metal reactor

    International Nuclear Information System (INIS)

    This safety evaluation report (SER) presents the final results of a preapplication design review for the Sodium Advanced Fast Reactor (SAFR) liquid metal reactor (Project 673). The SAFR conceptual design was submitted by the US Department of Energy (DOE) in accordance with the US Nuclear Regulatory Commission (NRC) ''Statement of Policy for the Regulation of Advanced Nuclear Power Plants'' (51 FR 24643 which provides for the early Commission review and interaction). The standard SAFR plant design consists of four identical reactor modules, referred to as ''paks,'' each with a thermal output rating of 900 MWt, coupled with four steam turbine-generator sets. The total electrical output was held to be 1400 MWe. This SER represents the NRC staff's preliminary technical evaluation of the safety features in the SAFR design. It must be recognized that final conclusions in all matters discussed in this SER require approval by the Commission. During the NRC staff review of the SAFR conceptual design, DOE terminated work on this design in September 1988. This SER documents the work done to that date and no additional work is planned for the SAFR

  10. An evolutionary approach to advanced water cooled reactors

    International Nuclear Information System (INIS)

    Based on the result of the Feasibility Study undertaken since 1991, Indonesia may enter in the new nuclear era by introduction of several Nuclear Power Plants in our energy supply system. Requirements for the future NPP's are developed in two step approach. First step is for the immediate future that is the next 50 years where the system will be dominated by A-LWR's/A-PHWR's and the second step is for the time period beyond 50 years in which new reactor systems may start to dominate. The integral reactor concept provides a revolutionary improvements in terms of conceptual and safety. However, it creates a new set of complex machinery and operational problems of its own. The paper concerns with a brief description of nuclear technology status in Indonesia and a qualitative assessment of integral reactor concept. (author)

  11. Advanced in the neutron feedback ICF reactor concept

    International Nuclear Information System (INIS)

    Results are reviewed and updated from an earlier design study of a novel nuclear-pumped flashlamp laser (NP-FL) inertial fusion energy (IFE) power reactor based on the neutron feedback concept for IFE. This concept includes nuclear pumping of the laser flashlamp, a D-T seeded D-3He target and magnetic protection of the first wall of the reactor chamber coupled with direct conversion of deflected charged particles. Advantages include an increased overall plant efficiency due to improved energy coupling via neutron feedback, increased thermal-to-electric energy conversion efficiency, and lower neutron activation and waste. These factors are reflected in a driver energy of 5 MJ and a target gain of only 50 for a 53 % efficient 1000-MWe power plant operating at 6 Hz, novel components involved. However, they require further technological development. Consequently, the NP-FL plant appears to provide a very attractive 'second-generation' IFE reactor. (authors)

  12. Advanced Neutron Source reactor control and plant protection systems design

    International Nuclear Information System (INIS)

    This paper describes the reactor control and plant protection systems' conceptual design of the Advanced Neutron Source (ANS). The Plant Instrumentation, Control, and Data Systems and the Reactor Instrumentation and Control System of the ANS are planned as an integrated digital system with a hierarchical, distributed control structure of qualified redundant subsystems and a hybrid digital/analog protection system to achieve the necessary fast response for critical parameters. Data networks transfer information between systems for control, display, and recording. Protection is accomplished by the rapid insertion of negative reactivity with control rods or other reactivity mechanisms to shut down the fission process and reduce heat generation in the fuel. The shutdown system is designed for high functional reliability by use of conservative design features and a high degree of redundance and independence to guard against single failures. Two independent reactivity control systems of different design principles are provided, and each system has multiple independent rods or subsystems to provide appropriate margin for malfunctions such as stuck rods or other single failures. Each system is capable of maintaining the reactor in a cold shutdown condition independently of the functioning of the other system. A highly reliable, redundant channel control system is used not only to achieve high availability of the reactor, but also to reduce challenges to the protection system by maintaining important plant parameters within appropriate limits. The control system has a number of contingency features to maintain acceptable, off-normal conditions in spite of limited control or plant component failures thereby further reducing protection system challenges

  13. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    International Nuclear Information System (INIS)

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

  14. Design study on sodium-cooled large-scale reactor

    International Nuclear Information System (INIS)

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

  15. Problems encountered and approaches for the deployment of advanced reactors in Eastern Europe

    International Nuclear Information System (INIS)

    Past developments in Eastern Europe resulted in a relatively high level of development of nuclear power in some countries. A brief study of the main features of this development and of the present situation in this region leads to the major conclusion that successful upgrading (or shutdown) of existing reactors (operational or under construction) is a necessary condition for public acceptance and at the same time is an essential pre-condition for the future development of nuclear power in Eastern Europe. Analysis of the actual situation in Bulgaria, the Czech Republic, the Russian Federation and Ukraine highlights the common problems and also the conflicting views. A set of necessary conditions which are of importance to all countries to permit the future deployment of advanced reactors is outline in the paper. These conditions are divided into three groups: general, technical and public acceptance conditions. On the basis of these conditions, the highest priority has to be allocated to the positive economic development of, and to the resulting need for, additional electricity generating sources. The second basic condition is the existence of a market for advanced reactors. The greatest possibility of entering such a worldwide market will be limited to countries with a stable and positive economy, like the Czech Republic, Hungary or Slovakia. The Russian Federation will probably continue isolated development of advanced reactor systems for its own needs. (author). 5 refs, 1 tab

  16. Advanced nuclear systems. Review study

    International Nuclear Information System (INIS)

    The task of this review study is to from provide an overview of the developments in the field of the various advanced nuclear systems, and to create the basis for more comprehensive studies of technology assessment. In an overview the concepts for advanced nuclear systems pursued worldwide are subdivided into eight subgroups. A coarse examination raster (set pattern) is developed to enable a detailed examination of the selected systems. In addition to a focus on enhanced safety features, further aspects are also taken into consideration, like the lowering of the proliferation risk, the enhancement of the economic competitiveness of the facilities and new usage possibilities (for instance concerning the relaxation of the waste disposal problem or the usage of alternative fuels to uranium). The question about the expected time span for realization and the discussion about the obstacles on the way to a commercially usable reactor also play a substantial role as well as disposal requirements as far as they can be presently recognized. In the central chapter of this study, the documentation of the representatively selected concepts is evaluated as well as existing technology assessment studies and expert opinions. In a few cases where this appears to be necessary, according technical literature, further policy advisory reports, expert statements as well as other relevant sources are taken into account. Contradictions, different assessments and dissents in the literature as well as a few unsettled questions are thus indicated. The potential of advanced nuclear systems with respect to economical and societal as well as environmental objectives cannot exclusively be measured by the corresponding intrinsic or in comparison remarkable technical improvements. The acceptability of novel or improved systems in nuclear technology will have to be judged by their convincing solutions for the crucial questions of safety, nuclear waste and risk of proliferation of nuclear weapons

  17. Design and layout decisions for refuelling system of advanced fast neutron reactor

    International Nuclear Information System (INIS)

    The experience in operation of BOR-60, BN-350 and BN-600 power units, as well as development of refuelling systems for BN-800 power unit, allows developing of refuelling system for BN-1200 advanced reactor of new generation. The refuelling system was developed on the basis of possible technical decisions aimed at improvement of safety and technical-and-economic indices. Structural layout of BN-1200 reactor refuelling system is given. Main differences in BN-1200 reactor refuelling system as compared with BN-800 reactor are given. Design features of refuelling equipment are: - BN-1200 reactor has a split large rotating plug to allow transporting of its components by railway with subsequent assembling at site; - the refuelling box is fabricated in the form of sectional parallelepiped to allow transporting of its components by railway with subsequent assembling at site; - one 'direct' refuelling mechanism and one cantilever' refuelling mechanism are used to refuel rarely replaced protection assemblies that allows reducing of overall dimensions of rotating plugs; - the vertical elevator is arranged on the oval plug installed on the reactor cover. The upper structure with elevator drive rotates together with the elevator plug under rotary drive located on the oval plug. The vertical elevator allows sufficient reduction of refuelling box; - the refuelling machine runs on straight-line rails. The vertical elevator, gas gate valve on reactor refuelling channel, non-use of spent FA drum and enhanced radiation protection on the column of refuelling box machine allows reduction of specific materials consumption of BN-1200 reactor refuelling system by more than 10 times as compared with BN-800 reactor. To verify refuelling equipment operability the following experiments are planned: - mastering of gripper design for 'direct' refuelling mechanism and refuelling machine; - mastering of 'cantilever' for refuelling mechanism; - mastering of fresh FA conveyor design. As for the

  18. Advanced drilling systems study.

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, Kenneth G.; Livesay, Billy Joe; Finger, John Travis (Livesay Consultants, Encintas, CA)

    1996-05-01

    This report documents the results of a study of advanced drilling concepts conducted jointly for the Natural Gas Technology Branch and the Geothermal Division of the U.S. Department of Energy. A number of alternative rock cutting concepts and drilling systems are examined. The systems cover the range from current technology, through ongoing efforts in drilling research, to highly speculative concepts. Cutting mechanisms that induce stress mechanically, hydraulically, and thermally are included. All functions necessary to drill and case a well are considered. Capital and operating costs are estimated and performance requirements, based on comparisons of the costs for alternative systems to conventional drilling technology, are developed. A number of problems common to several alternatives and to current technology are identified and discussed.

  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. Overview of the US program of controls for advanced reactors

    International Nuclear Information System (INIS)

    An automated control system can incorporate control goals and strategies, assessment of present and future plant status, diagnostic evaluation and maintenance planning, and signal and command validation. It has not been feasible to employ these capabilities in conventional hard-wired, analog, control systems. Recent advances in computer-based digital data acquisition systems, process controllers, fiber-optic signal transmission artificial intelligence tools and methods, and small inexpensive, fast, large-capacity computers---with both numeric and symbolic capabilities---have provided many of the necessary ingredients for developing large, practical automated control systems. Furthermore, recent reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. This paper presents an overall US national perspective for advanced controls research and development. The goals of high reliability, low operating cost and simple operation are described. The staged approach from conceptualization through implementation is discussed. Then the paper describes the work being done by ORNL, ANL and GE. The relationship of this work to the US commercial industry is also discussed

  1. Status of advanced containment systems for next generation water reactors

    International Nuclear Information System (INIS)

    The present IAEA status report is intended to provide information on the current status and development of containment systems of the next generation reactors for electricity production and, particularly, to highlight features which may be considered advanced, i.e. which present improved performance with evolutionary or innovative design solutions or new design approaches. The objectives of the present status report are: To present, on a concise and consistent basis, selected containment designs currently being developed in the world; to review and compare new approaches to the design bases for the containments, in order to identify common trends, that may eventually lead to greater worldwide consensus, to identify, list and compare existing design objectives for advanced containments, related to safety, availability, maintainability, plant life, decommissioning, economics, etc.; to describe the general approaches adopted in different advanced containments to cope with various identified challenges, both those included in the current design bases and those related to new events considered in the design; to briefly identify recent achievements and future needs for new or improved computer codes, standards, experimental research, prototype testing, etc. related to containment systems; to describe the outstanding features of some containments or specific solutions proposed by different parties and which are generally interesting to the international scientific community. 36 refs, 27 figs, 1 tab

  2. Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; Steven Prescott; Tony Koonce

    2014-04-01

    A key area of the Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) strategy is the development of methodologies and tools that will be used to predict the safety, security, safeguards, performance, and deployment viability of SMRs. The goal of the SMR PRA activity will be to develop quantitative methods and tools and the associated analysis framework for assessing a variety of risks. Development and implementation of SMR-focused safety assessment methods may require new analytic methods or adaptation of traditional methods to the advanced design and operational features of SMRs. We will need to move beyond the current limitations such as static, logic-based models in order to provide more integrated, scenario-based models based upon predictive modeling which are tied to causal factors. The development of SMR-specific safety models for margin determination will provide a safety case that describes potential accidents, design options (including postulated controls), and supports licensing activities by providing a technical basis for the safety envelope. This report documents the progress that was made to implement the PRA framework, specifically by way of demonstration of an advanced 3D approach to representing, quantifying and understanding flooding risks to a nuclear power plant.

  3. ADVANCED CUTTINGS TRANSPORT STUDY

    Energy Technology Data Exchange (ETDEWEB)

    Troy Reed; Stefan Miska; Nicholas Takach; Kaveh Ashenayi; Mark Pickell; Len Volk; Mike Volk; Evren Ozbayoglu; Lei Zhou

    2002-04-30

    This is the third quarterly progress report for Year 3 of the ACTS Project. It includes a review of progress made in: (1) Flow Loop construction and development and (2) research tasks during the period of time between Jan. 1, 2002 and Mar. 31, 2002. This report presents a review of progress on the following specific tasks: (a) Design and development of an Advanced Cuttings Transport Facility (Task 3: Addition of a Cuttings Injection/Separation System), (b) Research project (Task 6): ''Study of Cuttings Transport with Foam Under LPAT Conditions (Joint Project with TUDRP)'', (c) Research project (Task 9b): ''Study of Foam Flow Behavior Under EPET Conditions'', (d) Research project (Task 10): ''Study of Cuttings Transport with Aerated Mud Under Elevated Pressure and Temperature Conditions'', (e) Research on three instrumentation tasks to measure: Cuttings concentration and distribution in a flowing slurry (Task 11), Foam texture while transporting cuttings. (Task 12), and Viscosity of Foam under EPET (Task 9b); (f) Development of a Safety program for the ACTS Flow Loop, progress on a comprehensive safety review of all flow-loop components and operational procedures. (Task 1S); and (g) Activities towards technology transfer and developing contacts with Petroleum and service company members, and increasing the number of JIP members.

  4. ADVANCED CUTTINGS TRANSPORT STUDY

    Energy Technology Data Exchange (ETDEWEB)

    Troy Reed; Stefan Miska; Nicholas Takach; Kaveh Ashenayi; Gerald Kane; Mark Pickell; Len Volk; Mike Volk; Affonso Lourenco; Evren Ozbayoglu; Lei Zhou

    2002-01-30

    This is the second quarterly progress report for Year 3 of the ACTS project. It includes a review of progress made in: (1) Flow Loop development and (2) research tasks during the period of time between Oct 1, 2001 and Dec. 31, 2001. This report presents a review of progress on the following specific tasks: (a) Design and development of an Advanced Cuttings Transport Facility (Task 3: Addition of a Cuttings Injection/Collection System), (b) Research project (Task 6): ''Study of Cuttings Transport with Foam Under LPAT Conditions (Joint Project with TUDRP)'', (c) Research project (Task 9): ''Study of Foam Flow Behavior Under EPET Conditions'', (d) Research project (Task 10): ''Study of Cuttings Transport with Aerated Mud Under Elevated Pressure and Temperature Conditions'', (e) Research on instrumentation tasks to measure: Cuttings concentration and distribution in a flowing slurry (Task 11), and Foam properties while transporting cuttings. (Task 12), (f) Development of a Safety program for the ACTS Flow Loop. Progress on a comprehensive safety review of all flow-loop components and operational procedures. (Task 1S). (g) Activities towards technology transfer and developing contacts with Petroleum and service company members, and increasing the number of JIP members.

  5. Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

    Energy Technology Data Exchange (ETDEWEB)

    Wood, RT

    2004-09-27

    This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

  6. 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)

  7. Flow blockage analysis for the advanced neutron source reactor

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) reactor was designed to provide a research tool with capabilities beyond those of any existing reactors. One portion of its state-of-the-art design required high-speed fluid flow through narrow channels between the fuel plates in the core. Experience with previous reactors has shown that fuel plate damage can occur when debris becomes lodged at the entrance to these channels. Such debris disrupts the fluid flow to the plate surfaces and can prevent adequate cooling of the fuel. Preliminary ANS designs addressed this issue by providing an unheated entrance length for each fuel plate so that any flow disruption would recover, thus providing adequate heat removal from the downstream, heated portions of the fuel plates. As part of the safety analysis, the adequacy of this unheated entrance length was assessed using both analytical models and experimental measurements. The Flow Blockage Test Facility (FBTF) was designed and built to conduct experiments in an environment closely matching the ANS channel geometry. The FBTF permitted careful measurements of both heat transfer and hydraulic parameters. In addition to these experimental efforts, a thin, rectangular channel was modeled using the Fluent computational fluid dynamics computer code. The numerical results were compared with the experimental data to benchmark the hydrodynamics of the model. After this comparison, the model was extended to include those elements of the safety analysis that were difficult to measure experimentally. These elements included the high wall heat flux pattern and variable fluid properties. The results were used to determine the relationship between potential blockage sizes and the unheated entrance length required

  8. Coolant chemistry of the advanced carbon dioxide cooled reactor

    International Nuclear Information System (INIS)

    The large scale production of electricity by uranium fission has been achieved in the United Kingdom exclusively by reactors which are gas cooled and moderated by graphite. In this way the use of uranium with close to the natural isotopic content was possible. Once the choice of graphite as moderator had been made then the selection of a suitable gas to transport heat from the core to the steam generating equipment was limited and, in fact, only two have been identified as having suitable chemical and nuclear properties, namely helium and carbon dioxide. The first of these has the disadvantage of being expensive but has a high heat transfer capability and is fundamentally inert, its reactivity being controlled entirely by the level of impurities such as hydrogen and water. With the closure of the OECD Dragon High Temperature Reactor Project interest in helium cooling for nuclear plant has faded in the UK. The alternative coolant, carbon dioxide, which is cheap but chemically reactive is used in the first generation Magnox power stations and in the Commercial Advanced Gas Cooled Reactor (CAGR) design. In the more highly rated CAGR design the fuel consists of enriched uranium in the form of dioxide encased in stainless steel; the gas outlet temperature of the core is increased to around 6000C. The purpose of this paper is to provide an insight into the current thinking about the nature of the complex chemistry associated with the CAGR coolant and how this chemistry influences the rate of deposition onto the fuel pin surfaces and the rate of graphite moderator oxidation. The application of these ideas to the prediction of the behaviour of a CAGR core with particular reference to the calculation of coolant composition within the porous moderator structure at points remote from the surface, is outlined and the use of all this information to define a satisfactory range of coolant composition is also described

  9. ANDREA: Advanced nodal diffusion code for reactor analysis

    International Nuclear Information System (INIS)

    A new macro code is being developed at NRI which will allow coupling of the advanced thermal-hydraulics model with neutronics calculations as well as efficient use in core loading pattern optimization process. This paper describes the current stage of the macro code development. The core simulator is based on the nodal expansion method, Helios lattice code is used for few group libraries preparation. Standard features such as pin wise power reconstruction and feedback iterations on critical control rod position, boron concentration and reactor power are implemented. A special attention is paid to the system and code modularity in order to enable flexible and easy implementation of new features in future. Precision of the methods used in the macro code has been verified on available benchmarks. Testing against Temelin PWR operational data is under way (Authors)

  10. Computer code qualification program for the Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Ltd (AECL) has developed and implemented a Software Quality Assurance program (SQA) to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper provides an overview of the computer programs used in Advanced CANDU Reactor (ACR) safety analysis, and assessment of their applicability in the safety analyses of the ACR design. An outline of the incremental validation program, and an overview of the experimental program in support of the code validation are also presented. An outline of the SQA program used to qualify these computer codes is also briefly presented. To provide context to the differences in the SQA with respect to current CANDUs, the paper also provides an overview of the ACR design features that have an impact on the computer code qualification. (author)

  11. Advanced nuclear reactor public opinion project. Interim report

    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.

  12. Communication of Advanced Test Reactor probabilistic risk assessment results

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) probabilistic risk assessment (PRA) Level 1 results were efficiently communicated in two reports following the completion of revision 1 of the ATR PRA. As the ATR PRA including external events fills four large volume, it was considered impractical to expect all of the individuals at ATR who could benefit from the information to read the entire PRA. Even though many ATR personnel received training in PRA methodology and were involved in some aspects of the PRA, another hinderance to effective communication of the PRA results is that the PRA was written and organized to meet the requirements of practitioners and reviewers who are well-versed in PRA methods. Therefore, two PRA summary reports, an ATR risk summary report and an ATR functional group summary report, were written to communicate the ATR PRA results and insights to interested ATR personnel

  13. ADVANCED CUTTINGS TRANSPORT STUDY

    Energy Technology Data Exchange (ETDEWEB)

    Troy Reed; Stefan Miska; Nicholas Takach; Kaveh Ashenayi; Gerald Kane; Mark Pickell; Len Volk; Mike Volk; Barkim Demirdal; Affonso Lourenco; Evren Ozbayoglu; Paco Vieira; Lei Zhou

    2000-01-30

    This is the second quarterly progress report for Year 2 of the ACTS project. It includes a review of progress made in Flow Loop development and research during the period of time between Oct 1, 2000 and December 31, 2000. This report presents a review of progress on the following specific tasks: (a) Design and development of an Advanced Cuttings Transport Facility (Task 2: Addition of a foam generation and breaker system), (b) Research project (Task 6): ''Study of Cuttings Transport with Foam Under LPAT Conditions (Joint Project with TUDRP)'', (c) Research project (Task 7): ''Study of Cuttings Transport with Aerated Muds Under LPAT Conditions (Joint Project with TUDRP)'', (d) Research project (Task 8): ''Study of Flow of Synthetic Drilling Fluids Under Elevated Pressure and Temperature Conditions'', (e) Research project (Task 9): ''Study of Foam Flow Behavior Under EPET Conditions'', (f) Research project (Task 10): ''Study of Cuttings Transport with Aerated Mud Under Elevated Pressure and Temperature Conditions'', (g) Research on instrumentation tasks to measure: Cuttings concentration and distribution in a flowing slurry (Task 11), and Foam properties while transporting cuttings. (Task 12), (h) Development of a Safety program for the ACTS Flow Loop. Progress on a comprehensive safety review of all flow-loop components and operational procedures. (Task 1S). (i) Activities towards technology transfer and developing contacts with Petroleum and service company members, and increasing the number of JIP members. The tasks Completed During This Quarter are Task 7 and Task 8.

  14. ADVANCED CUTTINGS TRANSPORT STUDY

    Energy Technology Data Exchange (ETDEWEB)

    Troy Reed; Stefan Miska; Nicholas Takach; Kaveh Ashenayi; Gerald Kane; Mark Pickell; Len Volk; Mike Volk; Barkim Demirdal; Affonso Lourenco; Evren Ozbayoglu; Paco Vieira

    2000-10-30

    This is the first quarterly progress report for Year 2 of the ACTS project. It includes a review of progress made in Flow Loop development and research during the period of time between July 14, 2000 and September 30, 2000. This report presents information on the following specific tasks: (a) Progress in Advanced Cuttings Transport Facility design and development (Task 2), (b) Progress on research project (Task 8): ''Study of Flow of Synthetic Drilling Fluids Under Elevated Pressure and Temperature Conditions'', (c) Progress on research project (Task 6): ''Study of Cuttings Transport with Foam Under LPAT Conditions (Joint Project with TUDRP)'', (d) Progress on research project (Task 7): ''Study of Cuttings Transport with Aerated Muds Under LPAT Conditions (Joint Project with TUDRP)'', (e) Progress on research project (Task 9): ''Study of Foam Flow Behavior Under EPET Conditions'', (f) Initiate research on project (Task 10): ''Study of Cuttings Transport with Aerated Mud Under Elevated Pressure and Temperature Conditions'', (g) Progress on instrumentation tasks to measure: Cuttings concentration and distribution (Tasks 11), and Foam properties (Task 12), (h) Initiate a comprehensive safety review of all flow-loop components and operational procedures. Since the previous Task 1 has been completed, we will now designate this new task as: (Task 1S). (i) Activities towards technology transfer and developing contacts with Petroleum and service company members, and increasing the number of JIP members.

  15. Design features facilitating the decommissioning of advanced gas-cooled reactors

    International Nuclear Information System (INIS)

    The design of the advanced gas-cooled reactors is discussed as is the proposed decommissioning plan for delayed decommissioning. The special features which assist in decommissioning are presented. As a result of the study a catalogue of design features which will facilitate decommissioning is given. In addition to the catalogue of design features, the radioactive inventory 10 years after shutdown and 100 years after shutdown has been calculated. From this a provisional operator dose from activities associated with decommissioning has been assessed

  16. Progress in Helias reactor studies

    International Nuclear Information System (INIS)

    The Helias reactor is an upgraded version of the Wendelstein 7-X experiment, which is under construction in the city of Greifswald. The modular coil system comprises 50 coils, which are constructed using NbTi-superconducting cables. The basic dimensions are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 5 T, maximum field on coils 10 T. Over the past year progress toward better understanding of the fusion plasma has been made. In particular, the following issues have been addressed: Plasma equilibrium and MHD-stability; Neoclassical transport in the Helias configuration; Start-up scenarios and steady state burn; Alpha-particle orbits and alpha-particle losses; Drift waves in Helias configurations; Modelling the fusion plasma using empirical scaling laws Technical studies have been focussed on the optimization of the coil system with respect to magnetic field distribution, forces and stresses. In this context the ANSYS-code has been found useful for optimising the support system. In a first survey several blanket concepts, developed for the DEMO tokamak, have been adapted to the Helias geometry. Presently a water-cooled LiPb-blanket is favored in comparison with ceramic breeders, since safety properties and maintenance procedure seem to be more advantageous within this concept. Maintenance and replacement of blanket segments through portholes have also been studied with respect to their geometric compatibility. Finally parameter studies of low aspect ratio Helias reactors will be discussed. (orig.)

  17. Advanced neutron source reactor probabilistic flow blockage assessment

    International Nuclear Information System (INIS)

    The Phase I Level I Probabilistic Risk Assessment (PRA) of the conceptual design of the Advanced Neutron Source (ANS) Reactor identified core flow blockage as the most likely internal event leading to fuel damage. The flow blockage event frequency used in the original ANS PRA was based primarily on the flow blockage work done for the High Flux Isotope Reactor (HFIR) PRA. This report examines potential flow blockage scenarios and calculates an estimate of the likelihood of debris-induced fuel damage. The bulk of the report is based specifically on the conceptual design of ANS with a 93%-enriched, two-element core; insights to the impact of the proposed three-element core are examined in Sect. 5. In addition to providing a probability (uncertainty) distribution for the likelihood of core flow blockage, this ongoing effort will serve to indicate potential areas of concern to be focused on in the preliminary design for elimination or mitigation. It will also serve as a loose-parts management tool

  18. Seawater desalination using an advanced small integral reactor - SMART

    International Nuclear Information System (INIS)

    A concept of a dual-purpose integrated nuclear desalination plant coupled with the advanced small integral reactor SMART was established. The design concept of the plant aims to produce 40,000m5/day of water with the MED process and to generate about 90 MWe of electricity. In order to examine the technical, economic, and safety considerations in coupling SMART with desalination, a preliminary analysis on water production costs and a safety review of potential disturbances of the integrated nuclear desalination plant have been performed. The results of economic evaluation show that the use of SMART for seawater desalination is either comparative to or more economical, with respect to the water production cost, than the use of fossil fuels in comparison with the data published by the IAEA. It was also found that any possible transient event of the desalination plant does not impact on the reactor safety. The key safety parameters of the transient events induced by the potential disturbances of the desalination plant are bounded by the limits of safety analysis of SMART

  19. Reactor installation and maintenance for the Advanced Neutron Source

    International Nuclear Information System (INIS)

    Advanced Neutron Source (ANS) reactor assembly components have been modeled in great detail in IGRIP in order to realistically simulate preliminary installation and maintenance processes. Animation of these processes has been captured in a 15-minute video with narration. Approximately 90% of the parts were initially translated from CADAM (a two-dimensional drawing package) to IGRIP and then revolved or extruded. IGRIP's IGES translator greatly reduced the time required to perform this operation. The interfacing of devices in the work cell has identified numerous design inconsistencies. Most of the modeled reactor components are devices with a single degree of freedom (DOF) however, some of the slanted experiments required 6 DOF so that they could be removed at an angle in order to clear the reflector vessel flanges. IGRIP's collision detection feature proved to be extremely helpful in determining interferences when removing the experiments. The combination of three-dimensional visualization and collision detection allows engineers to clearly and easily visualize potential design problems before the construction phase of the project

  20. Modelling containment passive safety systems in advanced water cooled reactors

    International Nuclear Information System (INIS)

    Most designs of advanced passive reactors incorporate Passive Containment Cooling Systems (PCCS) relying on steam condensation to cope with possible pressure increase that would result in the case of a postulated accident. As a consequence, experimental and analytical research programmes have been launched worldwide to investigate new configurations and conditions involved in these new scenarios. This paper summarises the major outcomes of the joint research of CIEMAT, UPV, and UW in developing predictive models to address anticipated conditions in the Simplified Boiling Water Reactors (CIEMAT-UPV) and in the AP600 (CIEMAT-UW). Even though both models share some of their fundamental characteristics (such as being mass/heat transfer analogy based), samples of their validation against independent databases illustrate their intrinsic differences in formulation according to the scenarios addressed by each one. Relative importance of condensate film or gas mixture velocity are discussed, and the effect of key factors such as noncondensable gas presence and pressure are stated. Experimental data from University of Berkeley (UCB) and from University of Wisconsin - Madison (UW) will be used to support comparisons and discussions held in the paper. In short, this work demonstrates that heat/mass transfer analogy-based models, particularly those relying on diffusion film modelling to account for noncondensable gas presence, are extremely useful in test interpretation and result in good agreement with reliable databases. (author)

  1. Plasma control issues for an advanced steady state tokamak reactor

    International Nuclear Information System (INIS)

    This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

  2. Cooling of concrete structure in advanced heavy water reactor

    International Nuclear Information System (INIS)

    Innovative nuclear power plants are being designed by incorporation of passive systems to the extent possible for enhancing the safety by elimination of active components. BARC has designed Advanced Heavy Water Reactor (AHWR) incorporating several passive systems to facilitate the fulfillment of safety functions of the reactor during normal operation, residual heat removal, emergency core cooling, confinement of radioactivity etc. In addition to these passive systems, an innovative passive technology is being developed to protect, the concrete structure in high temperature zone (V1-volume). Passive Concrete Cooling System (PConCS) uses the principle of natural circulation to provide cooling outside the insulation cabinet encompassing high temperature piping. Cooling water is circulated from overhead GDWP in cooling pipes fixed over corrugated plate on outer surface of insulation cabinet and maintains low temperature of concrete structure. Modular construction of insulation cabinet and cooling pipes external to the concrete surface simplifies the design, construction and refurbishment if required. The paper describes the details of passive technology for concrete cooling. (author)

  3. Potential applications of robotics in advanced liquid-metal reactors

    International Nuclear Information System (INIS)

    The advanced liquid-metal reactor (ALMR) design includes a range of robots and automation devices. They extend from stationary robots that are a part of the current design to more exotic concepts with mobile, autonomous units, which may become part of the design. Development of robotic application requirements is enhanced by using computer models of work spaces in three dimensions. The primary goals of the more autonomous machines are to: (1) extent and/or enhance one's capabilities in a hazardous environment; some tasks could encounter high temperatures (up to 800 degree F), high radiation (fields up to several hundred thousand roentgens per hour), rooms filled with inert gas and/or sodium aerosol, or combinations of these; (2) reduce operating and maintenance cost through inservice inspection (ISI) of various parts of the reactor, through consideration of as-low-as-reasonably achievable radiation levels, and through automation of some maintenance/processing operations. This paper discusses some applications in the fuel cycle, in refueling operations, and in inspection

  4. 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)

  5. Advanced SFR concept design studies at KAERI

    International Nuclear Information System (INIS)

    Full text: Advanced SFR design concepts have been proposed and evaluated against the design requirements to satisfy the Gen IV technology goals. Two types of conceptual core designs, Breakeven and TRU burner cores were developed. Breakeven core is 1,200 MWe and does not have blankets to enhance the proliferation resistance. According to the current study, TRU burning rate increases linearly with the rated core powers from 600 MWe to 1,200 MWe. Considering 1) the realistic size of an SFR demonstration reactor for the long-term R and D plan with the goal of a demonstration SFR construction by 2028, and 2) the availability of a KALIMER-600 reactor system design that was developed in the last R and D phase, a TRU burner of 600 MWe was selected. The heat transport system of Advanced SFR was designed to be a pool type to enhance system safety through slow system transients, where primary sodium is contained in a reactor vessel. The heat transport system is composed of Primary Heat Transport System (PHTS), Intermediate Heat Transport System (IHTS), Steam Generating System (SGS) and Residual Heat Removal System (RHRS). The heat transport system was established through trade studies in order to enhance the safety and to improve the economics and performance of the KALIMER-600 design. Trade studies were performed for the number of IHTS loops, the number of PHTS pumps, Steam Generator (SG) design concepts, energy conversion system concepts, cover gas operation methods, and an improved concept of safety-graded passive decay heat removal system. From the study, the heat transport system of Advanced SFR has design features such as two IHTS loops, a Rankine cycle energy conversion system, two double-wall straight tube type SGs, and a passive decay heat removal system. In order to secure the economic competitiveness of an SFR, several concepts were implemented in the mechanical structural design without losing the reactor safety level. The material of reactor vessel and internal

  6. Conceptual Design Study of JSFR (2) - Reactor System

    International Nuclear Information System (INIS)

    Several innovative technologies are adopted in the JSFR design to meet the high level requirements for economic competitiveness in the design requirements. The cost-down approaches for JSFR are as follows. In order to reduce the amount of structural materials, the diameter of the reactor vessel shall be minimized and the reactor internal structures shall be simplified. The reduction of the reactor vessel diameter is achieved by adopting a advanced refueling system and the hot reactor vessel with high temperature wall. The flow velocity in the reactor upper plenum increases because the diameter of the reactor vessel is decreased. As the result, the coolant flow field in reactor upper plenum is severe. The optimization of the coolant flow field in the reactor upper plenum was carried out for prevention the cover gas entrainment and the vortex cavitations at the hot leg intake. In addition, structural integrities for seismic loadings and thermal loadings were evaluated because the design seismic loading was highly increased and the vessel wall is directly exposed to the thermal transients of the upper plenum. This paper describes the characteristics and the results of the design study of the reactor system. (author)

  7. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    International Nuclear Information System (INIS)

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830

  8. Carbide and nitride fuels for advanced burner reactor

    International Nuclear Information System (INIS)

    Full text: Under the U.S. fast reactor program, reference and alternative 1000 MWth Advanced Burner Reactor (ABR) core concepts were developed using ternary metallic (U-TRU-Zr) and mixed oxide (UO2+TRUO2) fuels. Recently, mixed carbide and nitride fuels have been considered as fast reactor fuels on the basis of their high density, compatibility with coolant, high melting temperature, and excellent thermal conductivity although they are ceramic fuel like a mixed oxide fuel. Thus, the performance of the ABR core loaded with carbide and nitride fuels was evaluated in this study with an expectation that the carbide and nitride fuels can mitigate disadvantages of both metallic and oxide fuels in the ABR: favorable passive safety features in a severe accident compared to the oxide core, a higher discharge burnup compared to the metallic core, and a potential to increase thermal efficiency. All calculations performed in this study were focused on the neutronics characteristics, although the fabrication and irradiation experiences for carbide and nitride fuels are limited and some problems were observed in the reprocessing and irradiation of these fuels. The mixed monocarbide and mixed mononitride fuels were selected as the alternative fuel forms and the ABR core concepts with these fuels were developed based on the reference 1000 MWth ABR core concepts. For consistency, the potential design goals used in the reference ABR core concepts were also employed in this study: a 1000 MWth power rating, medium TRU conversion ratio of ∼0.75, a compact core, one-year operational cycle length at least with a capacity factor of 90%, sufficient shutdown margin with a limited maximum single control assembly fault, and possible use of either metallic or any ceramic fuels in the same core layout. The core layout and outer assembly dimensions of the reference 1000 MWth ABR core were kept, but the intra assembly design parameters were varied to maximize the discharge burnup within the

  9. Large-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    International Nuclear Information System (INIS)

    The PIUS advanced reactor is a 640-MWe pressurized water reactor concept developed by Asea Brown Boveri. A unique feature of PIUS is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. Los Alamos is supporting the US Nuclear Regulatory Commission's preapplication review of the PIUS reactor. Baseline calculations of the PIUS Supplement design were performed for a large-break loss-of-coolant (LBLOCA) initiator using TRAC-PF1/MOD2. Additional sensitivity studies examined flow blockage and boron dilution events to explore the robustness of the PIUS concept for low-probability combination events following an LBLOCA

  10. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2009-09-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). 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. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  11. Physics design of advanced heavy water reactor utilising thorium

    International Nuclear Information System (INIS)

    An Advanced Heavy Water Reactor (AHWR) is being developed in India with the aim of utilising thorium for power generation. AHWR is a vertical pressure tube type reactor cooled by boiling light water and moderated by heavy water. It has been optimised for the thorium cycle. The main design objective is to be self-sustaining in 233U with most of the power from the thorium fuel using plutonium as the external fissile feed. It incorporates several advanced safety features namely, heat removal through natural circulation and a negative void coefficient of reactivity. The reactor has been designed to produce 750 MW(th) at a discharge burnup of 20,000 MWd/H(e). The physics design of AHWR has followed an evolutionary path ranging from a seed and blanket concept to a simplified composite cluster to achieve a good thermal hydraulic coupling. We have designed a composite cluster using both kinds of fuel namely, (Th-UO2 and (Th-Pu)O2. With plutonium seed, negative void coefficient can be achieved by making the spectrum harder. This was done by using a pyrocarbon scatterer in the moderator. The void coefficient strongly depends on plutonium. As plutonium burns very rapidly, it is not possible to achieve uniformly negative void coefficient with burnup in this cluster. Alternatively, burnable poison can be used within the cluster to achieve negative void coefficient taking advantage of the flux redistribution and change in spectrum upon voiding. Here, it is possible to achieve almost constant void reactivity with burnup resulting in a good thermal hydraulic coupling. The cluster design presently incorporates a central burnable absorber region. Boiling light water coolant requires that the core power distribution be optimised with thermal hydraulic parameters. The peaking factors inside the cluster should be low so as to have significant margin in operational conditions and to avoid burnout in accident conditions. The variation of reactivity from cold clean to hot operating has

  12. A modern control room for Indian Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor (AHWR) is a next generation nuclear power plant being developed by Bhabha Atomic Research Centre, India. AHWR is a vertical, pressure tube type, heavy-water-moderated, boiling light-water-cooled, innovative reactor, relying on natural circulation for core cooling in all operating and accident conditions. In addition, it incorporates various passive systems for decay heat removal, containment cooling and isolation. In addition to the many passive safety features, AHWR has state of the art I and C architecture based on extensive use of computers and networking. In tune with the many advanced features of the reactor, a centralized modern control room has been conceived for operation and monitoring of the plant. The I and C architecture enables the implementation of a fully computerised operator friendly control room with soft Human Machine Interfaces (HMI). While doing so, safety has been given due consideration. The control and monitoring of AHWR systems are carried out from the fully computer-based operator interfaces, except safety systems, for which only monitoring is provided from soft HMI. The control of the safety systems is performed from dedicated hardwired safety system panels. Soft HMI reduces the number of individual control devices and improves their reliability. The paper briefly describes the I and C architecture adopted for the AHWR plant along with the interfaces to the main and backup control rooms. There are many issues involved while introducing soft HMI based operator interfaces for Nuclear Power Plants (NPP) compared to the conventional plants. Besides discussing the implementation issues, the paper elaborates the design considerations that have undergone in the design of various components in the main control room especially operator workstations, shift supervisor console, safety system panels and large display panels. Mainly task based displays have been adopted for the routine operator interactions of the plant

  13. Conceptual design study of fusion experimental reactor (FY86 FER)

    International Nuclear Information System (INIS)

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

  14. Modelling of advanced structural materials for GEN IV reactors

    International Nuclear Information System (INIS)

    The choice of suitable materials and the assessment of long-term materials damage are key issues that need to be addressed for the safe and reliable performance of nuclear power plants. Operating conditions such as high temperatures, irradiation and a corrosive environment degrade materials properties, posing the risk of very expensive or even catastrophic plant damage. Materials scientists are faced with the scientific challenge to determine the long-term damage evolution of materials under service exposure in advanced plants. A higher confidence in life-time assessments of these materials requires an understanding of the related physical phenomena on a range of scales from the microscopic level of single defect damage effects all the way up to macroscopic effects. To overcome lengthy and expensive trial-and-error experiments, the multiscale modelling of materials behaviour is a promising tool, bringing new insights into the fundamental understanding of basic mechanisms. This paper presents the multiscale modelling methodology which is taking root internationally to address the issues of advanced structural materials for Gen IV reactors

  15. Impact of cut-off probability on emergency planning zone sizing for advanced light water reactors

    International Nuclear Information System (INIS)

    In 1997, NRC performed an evaluation of emergency planning for the evolutionary and advanced light water reactors, and indicated that no changes to emergency planning (EP) requirements were warranted in the existing technical framework. How- ever, NRC also recognized if the lower probability of severe accidents or the longer time period between accident initiation and release of radioactive material was considered, the changes to EP requirements for advanced light water reactors might be warranted and the emergency planning zone (EPZ) sizing might be reduced. This means that if the accidents which probabilities are lower than a cut-off probability are not considered, there may be a significant impact on the EPZ sizing. AP1000 was selected as an example in this study and the probable impacts of different cut-off probabilities were studied by using MACCS code in accordance with the NUREG-0396 method. The results show that, only if the cut-off probability is higher than the occurrence probability of the accident which has relatively serious consequence and high probability, there will be a significant impact on the EPZ sizing for the advanced light water reactor. (authors)

  16. Computer simulation of homogenization of boric acid in a pressurizer of the advanced nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rosa, Jose E.P. da; Moreira, Maria de L., E-mail: jeduird@hotmail.com, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Oliveira, Andre F. de, E-mail: eafoliveira@ien.gov.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2013-07-01

    The reactivity of a water cooled reactor is controlled using control rods or boron dilution in water of the primary circuit. The boron-10 ({sup 10} B) is an efficient neutron absorber, especially when used in the absorption of thermal neutrons. Transient studies with disabilities in the homogenization of boron in PWR reactors become important as the boric acid solution is added to the primary circuit coolant in order to help control the fission rate in the reactor core. After reactor shutdown, the boron present in coolant has the function of maintaining reactor subcriticality. If low concentrated boron solution enters in the primary circuit, it becomes necessary to inject boron and to assure that the coolant will be well homogenized in order to increase the concentration and thus preventing water with small amounts of boron to reach the core. The aim of this study is to simulate the boron homogenization in the pressurizer of an advanced nuclear reactor. It is used a test section, which represents a quarter of a modular nuclear reactor pressurizer. By using the CFX code, a computer program that allows thermal hydraulic analysis of different types of flow, three examples were simulated using different operating conditions. With the results, it was analyzed the parameters that could influence this homogenization. Case studies such as variation of the dimensions of the water inlet and outlet tubes, flow variation and change in positioning of entrances and exits were made with the goal of finding parameters that could help the optimization of the homogenization of boron. The results confirm that the issues analyzed can be changed in the project in order to obtain the best operating condition. (author)

  17. Overview of the Consortium for the Advanced Simulation of Light Water Reactors (CASL)

    OpenAIRE

    Kulesza Joel A.; Franceschini Fausto; Evans Thomas M.; Gehin Jess C.

    2016-01-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) was established in July 2010 for the purpose of providing advanced modeling and simulation solutions for commercial nuclear reactors. The primary goal is to provide coupled, higher-fidelity, usable modeling and simulation capabilities than are currently available. These are needed to address light water reactor (LWR) operational and safety performance-defining phenomena that are not yet able to be fully modeled taking a fir...

  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. 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)

  20. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  1. Proceedings of the 1992 topical meeting on advances in reactor physics

    International Nuclear Information System (INIS)

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements ampersand Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  2. 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

  3. D-D tokamak reactor studies

    International Nuclear Information System (INIS)

    A tokamak D-D reactor design, utilizing the advantages of a deuterium-fueled reactor but with parameters not unnecessarily extended from existing D-T designs, is presented. Studies leading to the choice of a design and initial studies of the design are described. The studies are in the areas of plasma engineering, first-wall/blanket/shield design, magnet design, and tritium/fuel/vacuum requirements. Conclusions concerning D-D tokamak reactors are stated

  4. Accelerator-driven systems (ADS) and fast reactors (FR) in advanced nuclear fuel cycles - a comparative study. Extract from OECD/NEA Publication ISBN 92-64-184882-1

    International Nuclear Information System (INIS)

    The long-term hazard of radioactive wastes arising from nuclear energy production is a matter of continued discussion and public concern in many countries. By the use of partitioning and transmutation (P and T) of the actinides and some of the long-lived fission products, the radiotoxicity of the high-level waste (HLW) and, possibly, the safety requirements for its geological disposal can be reduced compared with the current once-through fuel cycle. To make the technologically complex enterprise worthwhile, a reduction in the HLW radiotoxicity by a factor of at least one hundred is desirable. This requires very effective reactor and fuel cycle strategies, including fast reactors (FRs) and/or accelerator-driven, sub-critical systems. The accelerator-driven system (ADS) has recently been receiving increased attention due to its potential to improve the flexibility and safety characteristics of transmutation systems. The present study compares FR- and ADS-based actinide transmutation systems with respect to reactor properties, fuel cycle requirements, economic aspects, and R and D needs. The essential differences between the various systems are evaluated with the help of a number of representative 'fuel cycle schemes'. The strategies investigated include an evolutionary transmutation strategy in which the ADS provides additional flexibility by enabling plutonium utilisation in conventional reactors and confining the minor actinides to a small part of the fuel cycle, and two innovative transuranics (TRU) burning strategies, with an FR or an ADS, in which plutonium and minor actinides are managed together to minimise the proliferation risk. A novelty in the present study is that the analyses are carried out in a consistent manner using reactor and fuel cycle parameters which have been agreed upon by international experts. Principle messages from the study which could influence P and T policy development are: - Fuel cycles with multiple recycling of the fuel

  5. Effects of Plant Capital Costs on Market Introduction of Advanced Reactors

    International Nuclear Information System (INIS)

    In order to make reasonable forecasts about the future market shares of the different reactor types in the energy market, scenario studies are used. For this purpose, the future nuclear reactor park mix in Europe has been analysed applying an integrated dynamic process modelling technique. Starting point in the analyses is the current nuclear reactor park in the EU27 countries, taken into account the foreseen lifetime of each individual reactor as well as the nuclear fuel cycle infrastructure. Furthermore, an energy demand scenario, which is derived from the World Economic Council and the European DG-TREN, has been applied as input for the analyses. Various market share scenarios for nuclear energy are derived including sub-variants with regard to the intra-nuclear options taken, e.g. introduction date of Gen-III (i.e. EPR) and Gen-IV (i.e. SCWR, HTR, FR) reactors, level of reprocessing, and so forth. Realistic assumptions with respect to introduction dates of advanced reactors, reactor characteristics and fuel cycle facility characteristics were taken attempt-ing to provide an as realistic as possible framework for the assessment of future nuclear energy system scenarios in Europe. The assessment was undertaken using the DANESS code (Dynamic Analysis of Nuclear Energy System Strategies, developed by Argonne National Laboratory) and providing a complete picture of mass-flow and economics of the various nuclear energy system scenarios. The present assessment recognizes the integrated nuclear fuel cycle and foresees the evolution in cost of electricity. For the evaluation of the different future development paths for nuclear energy, reactor, front- and back- end specific cost factors are associated with the mass flow of uranium, plutonium, and minor actinides. The energy costing model covers capital cost, O and M costs and fuel cycle costs and consequently calculates levelized cost of electricity for the given nuclear system. The analyses show that the future

  6. Advanced marine reactor MRX and its application for electricity and heat co-generation

    International Nuclear Information System (INIS)

    The basic concept of an innovative advanced marine reactor MRX has been established by design study toward the goals of light-weightiness, compactness, and safety and reliability improvement with adoption of several new technologies. The MRX is the integral-type PWR aimed for use of ship propulsion. Adoption of a water-filled containment makes the reactor light-weighted and compact greatly. The total weight and volume of the reactor are 1600 tons and 1210 m3, which are equivalent to halves of the Mutsu, although the reactor power of MRX is three times greater. An engineered safety system of the MRX is a simplified passive system, function of which is confirmed by the safety analysis to be able to keep the reactor integrity even in a case of accident. Reliability of the system is evaluated by the PSA and revealed to have two orders smaller core damage occurrence frequency than existing PWRs. The MRX can be applied to an energy supply system of electricity and heat co-generation. Concept of the nuclear energy supply system is designed to generate electricity, heat and fresh water. The nuclear barge is shown to be a possible nuclear energy supply system with advantage of being easily moveable. (author)

  7. An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE-ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

  8. Advanced nuclear fuel for VVER reactors. Status and operation experience

    International Nuclear Information System (INIS)

    The paper discusses the major VVER fuel trends, aimed at the enhancement of FAs' effectiveness and reliability, flexibility of their operating performances and fuel cycle efficiency, specifically: (i) Fuel burnup increasing is one of the major objectives during the development of improved nuclear fuel and fuel cycles. At present, the achieved fuel rod burn up is 65 MWdays/kgU. The tasks are set and the activities are carried out to achieve fuel rod burnup up to 70 MWdays/kgU and burnup of discharged batch of FAs - up to 60 MWdays/kgU. (ii) Improvement of FA rigidity enables to increase operating reliability of fuel due to gaps reducing between FAs and, as a result, the fall of peak load coefficients. FA geometric stability enables to optimize the speed of handling procedures with fuel. (iii) Increasing of uranium content of FA is aimed at extension of fuel cycles' duration. Fuel weight increase in FA is achieved both due to fuel column height extension and to changes of pellet geometrical size. (iv) Extension of FA service live satisfies the up-to-date NPP requirements for fuel cycles of various duration from 4x320 eff. days to 5x320 eff. days and 3x480 eff. days. (v) The development of new-generation FAs with increased strength characteristics has required the zirconium alloys' improvement. Advanced zirconium alloys shall provide safety and effectiveness of FA and fuel rods during long-life operation up to 40 000 eff. hours. (vi) Utilization of reprocessed uranium enables to use spent nuclear fuel in cycle and to create the partly complete fuel cycle for VVER reactors. This paper summarizes the major operating results of LTAs, which meet the modern and prospective requirements for VVER fuel, at Russian NPPs with VVER-440 and VVER-1000 reactors. (author)

  9. Update on ORNL TRANSFORM Tool: Simulating Multi-Module Advanced Reactor with End-to-End I&C

    Energy Technology Data Exchange (ETDEWEB)

    Hale, Richard Edward [ORNL; Fugate, David L [ORNL; Cetiner, Sacit M [ORNL; Qualls, A L [ORNL

    2015-05-01

    The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the fourth year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled reactor) concepts, including the use of multiple coupled reactors at a single site. The focus of this report is the development of a steam generator and drum system model that includes the complex dynamics of typical steam drum systems, the development of instrumentation and controls for the steam generator with drum system model, and the development of multi-reactor module models that reflect the full power reactor innovative small module design concept. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor models; ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface technical area; and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the TRANSFORM tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the Advanced Reactors Technology program; (2) developing a library of baseline component modules that can be assembled into full plant models using available geometry, design, and thermal-hydraulic data; (3) defining modeling conventions for interconnecting component models; and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

  10. Advanced CANDU reactor development: a customer-driven program

    International Nuclear Information System (INIS)

    The Advanced CANDU Reactor (ACR) product development program is well under way. The development approach for the ACR is to ensure that all activities supporting readiness for the first ACR project are carded out in parallel, as parts of an integrated whole. In this way design engineering, licensing, development and testing, supply chain planning, construct ability and module strategy, and planning for commissioning and operations, all work in synergy with one another. Careful schedule management :ensures that program focus stays on critical path priorities.'This paper provides an overview of the program, with an emphasis on integration to ensure maximum project readiness, This program management approach is important now that AECL is participating as the reactor vendor in Dominion Energy's DOE-sponsored Combined Construction/Operating License (COL) program. Dominion Energy selected the ACR-700 as their reference reactor technology for purposes of demonstrating the COL process. AECL's development of the ACR is unique in that pre-licensing activities are being carded out parallel in the USA and Canada, via independent, but well-communicated programs. In the short term, these programs are major drivers of ACR development. The ACR design approach has been to optimize to achieve major design objectives: capital cost reduction, robust design with ample margins, proveness by using evolutionary change from existing :reference plants, design for ease :of operability. The ACR development program maintains these design objectives for each of the program elements: Design: .Carefully selected design innovations based on the SEU fuel/light water coolant:/heavy water moderator approach. Emphasis on lessons-learned review from operating experience and customer feedback Licensing: .Safety case based on strengths of existing CANDU plus benefits of optimised design Development and Test: Choice of materials, conditions to enable incremental testing building on existing CANDU and LWR

  11. ADVANCED CUTTINGS TRANSPORT STUDY

    Energy Technology Data Exchange (ETDEWEB)

    Stefan Miska; Troy Reed; Ergun Kuru

    2004-09-30

    The Advanced Cuttings Transport Study (ACTS) was a 5-year JIP project undertaken at the University of Tulsa (TU). The project was sponsored by the U.S. Department of Energy (DOE) and JIP member companies. The objectives of the project were: (1) to develop and construct a new research facility that would allow three-phase (gas, liquid and cuttings) flow experiments under ambient and EPET (elevated pressure and temperature) conditions, and at different angle of inclinations and drill pipe rotation speeds; (2) to conduct experiments and develop a data base for the industry and academia; and (3) to develop mechanistic models for optimization of drilling hydraulics and cuttings transport. This project consisted of research studies, flow loop construction and instrumentation development. Following a one-year period for basic flow loop construction, a proposal was submitted by TU to the DOE for a five-year project that was organized in such a manner as to provide a logical progression of research experiments as well as additions to the basic flow loop. The flow loop additions and improvements included: (1) elevated temperature capability; (2) two-phase (gas and liquid, foam etc.) capability; (3) cuttings injection and removal system; (4) drill pipe rotation system; and (5) drilling section elevation system. In parallel with the flow loop construction, hydraulics and cuttings transport studies were preformed using drilling foams and aerated muds. In addition, hydraulics and rheology of synthetic drilling fluids were investigated. The studies were performed under ambient and EPET conditions. The effects of temperature and pressure on the hydraulics and cuttings transport were investigated. Mechanistic models were developed to predict frictional pressure loss and cuttings transport in horizontal and near-horizontal configurations. Model predictions were compared with the measured data. Predominantly, model predictions show satisfactory agreements with the measured data. As a

  12. The nuclear fuel cycle with advanced reactor systems - analysis of its economic fundamentals and possibilities

    International Nuclear Information System (INIS)

    The purpose of the study is to analyse the nuclear fuel cycle of alternative advanced reactor systems with respect to their different mass flows of nuclear fuel and to judge the economic feasibility of these advanced nuclear technologies using a specific fuel cycle model. It is the particular importance of this subject that many technical, physical, political and economic coherences are combined in a very complex manner. A detailed description of the problem is given in the introductional chapter 1. The following chapter 2 gives a sufficient survey of the different techniques and technical facilities of the nuclear fuel cycles in question. Part 3 includes an investigation of logical coherences between typical fuel cycle mass flows which consequently leads to a mathematical model. This model is described in part 4. Chapter 5 then deals with the application of this model by the quantitative estimation and valuation of the economic differences between the conventional and advanced nuclear technology. In the final part of this study the influence of a very important parameter in this context, the price of plutonium, is discussed with respect to the time of introduction of the advanced reactor technology under economic conditions. (orig.)

  13. 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)

  14. 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)

  15. Parametric Evaluation of Large-Scale High-Temperature Electrolysis Hydrogen Production Using Different Advanced Nuclear Reactor Heat Sources

    International Nuclear Information System (INIS)

    High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 C to 950 C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycles producing the highest efficiencies varied depending on the temperature range considered

  16. Parametric evaluation of large-scale high-temperature electrolysis hydrogen production using different advanced nuclear reactor heat sources

    International Nuclear Information System (INIS)

    High-temperature electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800-950 oC, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an intermediate heat exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed with the objective of evaluating the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency of the integrated plant design for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycles producing the highest efficiencies varied depending on the temperature range considered.

  17. Advances in shielding calculations for the PEC reactor

    International Nuclear Information System (INIS)

    In this paper calculations of neutron and gamma streaming through various penetrations in the plug and neutron shield of the sodium cooled fast reactor PEC, currently under construction, are described. The object of the calculations has been to verify the accessibility, 3 days after reactor shut-down, of the area directly above the reactor. (author)

  18. PARs for combustible gas control in advanced light water reactors

    International Nuclear Information System (INIS)

    This paper discusses the progress being made in the United States to introduce passive autocatalytic recombiner (PAR) technology as a cost-effective alternative to electric recombiners for controlling combustible gas produced in postulated accidents in both future Advanced Light Water Reactors (ALWRs) and certain U. S. operating nuclear plants. PARs catalytically recombine hydrogen and oxygen, gradually producing heat and water vapor. They have no moving parts and are self-starting and self-feeding, even under relatively cold and wet containment conditions. Buoyancy of the hot gases they create sets up natural convective flow that promotes mixing of combustible gases in a containment. In a non-inerted ALWR containment, two approaches each employing a combination of PARs and igniters are being considered to control hydrogen in design basis and severe accidents. In pre-inerted ALWRs, PARs alone control radiolytic oxygen produced in either accident type. The paper also discusses regulatory feedback regarding these combustible gas control approaches and describes a test program being conducted by the Electric Power Research Institute (EPRI) and Electricite de France (EdF) to supplement the existing PAR test database with performance data under conditions of interest to U.S. plants. Preliminary findings from the EPRI/EdF PAR model test program are included. Successful completion of this test program and confirmatory tests being sponsored by the U. S. NRC are expected to pave the way for use of PARs in ALWRs and operating plants. (author)

  19. Carbide and Nitride Fuels for Advanced Burner Reactor

    International Nuclear Information System (INIS)

    The impacts of the mixed carbide and nitride fuels on the core performances and passive safety features of TRU burner were assessed and comapred with the metallic and oxide fuels. Targeting the potential design goals adopted in the Advanced Burner Reactor core concepts, the alternative TRU burner concepts were developed by loading carbide and nitride fuels. The neutron spectrum is softer than that of the metal core, but harder than that of the oxide core, and the core performance parameters such as fuel residence time, discharge burnup, flux level, etc are generally between the values of the metal and oxide cores. The margin to fuel melt was significantly increased because of the high thermal conductivity and high melting temperature, and hence there is an additional room to improve the thermal efficiency by increasing the operating temperature. The changed fuel composition affected the kinetics parameters and reactivity feedback coefficients, but the variations were minimal. The reduced core height decreases the sodium void worth, and the high thermal conductivity decreases the fuel temperature and Doppler constant. As a result, both carbide and nitride cores have favorable passive safety features without additional design fixes that are required in the oxide core concepts. (author)

  20. Overview of human factors issues associated with advanced passive reactors

    International Nuclear Information System (INIS)

    Nuclear power plant operators in current plants do just that, they operate the plant. They monitor the state of the process, observe trends, react to system changes and take action to maintain the plant in a safe condition. In order to accomplish these activities, they (1) rely on information available to them from the various instruments and displays in the control room, (2) use hard copy procedures to direct their actions and (3) perform their control actions through the various switches, keyboards and knobs available. Proposals from vendors and observations of activities in other countries indicate that the control room for advanced passive reactors is likely to be quite different from control rooms one would seen in current power plants. These differences will go beyond the immediate appearances, e.g., smaller, more cockpit-like, to issues of automation, function-allocation, the changing role of the operators and their training and qualifications, and software verification and validation. This research paper discusses some of these planned efforts and how they relate to the human factors issues and regulatory user needs

  1. 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.)

  2. Utility Leadership in Defining Requirements for Advanced Light Water Reactors

    International Nuclear Information System (INIS)

    It is appropriate, based on twenty five years of operating experience, that utilities take a position of leadership in developing the technical design and performance requirements for the next generations of nuclear electric generating plants. The U. S. utilities, through the Electric Power Research Institute, began an initiative in 1985 to develop such Utility requirements. Many international Utility organizations, including Korea Electric Power Corporation, have joined as full participants in this important Utility industry initiative. In light of the closer linkage among countries of the world due to rapid travel and telecommunications, it is also appropriate that there be international dialogue and agreement on the principal standards for nuclear power plant acceptability and performance. The Utility/EPRI Advanced Light Water Reactor Program guided by the ALRR Utility Steering Committee has been very successful in developing these Utility requirements. This paper will summarize the state of development of the ALRR Utility Requirements for Evolutionary Plants, recent developments in their review by the U. S. Nuclear Regulatory Commission, resolution of open issues, and the extension of this effort to develop a companion set of ALRR Utility Requirements for plants employing passive safety features

  3. On the path to ordering standardized advanced light water reactors

    International Nuclear Information System (INIS)

    The international Advanced Light Water Reactor (ALWR) program is specifying, designing, and certifying the next generation of nuclear power plants. Begun in the mid-1980's, the program is on track to permit ordering and construction of families of standardized plants at the start of the twenty-first century. ALWRs will be constructed only if they are economically competitive with alternative forms of electricity generation and are recognized as acceptable and favorable by the public, prospective owners, and investors. This paper first gives an overview of the major building blocks ensuring safe, reliable, and economic designs and the status of those designs. Next it lays out the path the industry has charted toward adopting the ALWR option and indicates the status of three key steps -- design certification, utility requirements, and first-of-a-kind engineering. Lastly, the paper focuses on one of the most important building blocks for ensuring economic viability -- life-cycle standardization. Among the topics are the definition and scope of standardization; its advantages and disadvantages; design team standardization plans that describe the desired or optimum degree of standardization and the processes used to achieve it; and the need for an agreement among all plant owners and operators for implementing and sustaining standardization in families of ALWRs. 10 refs., 5 figs

  4. Advanced CANDU reactor technology: competitive design for the nuclear renaissance

    International Nuclear Information System (INIS)

    AECL has developed the design for a new generation of CANDU nuclear power plants, the Advance CANDU Reactor or ACR. The ACR combines a set of underlying enabling technologies with well-established successful CANDU features in an optimized design with significantly lower costs. By adopting slightly enriched uranium fuel, an optimized core design with light water coolant, heavy water moderator and reflector has been defined based on the existing CANDU fuel channel module. The basic design for the complete reference ACR power plant has now been completed. This paper summarizes the main features and characteristics of the reference ACR-700 power plant design. The progress of the ACR design program in meeting challenging cost, schedule and performance targets is described. AECL's cost reduction methodology is summarized as an integral part of the design optimization process. Examples are given of cost reduction features together with the enhancement of design margins. AECL expects the detailed design and testing of ACR to be complete and pre-project licensing evaluation carried out to enable regulatory endorsement in key markets by the middle of the decade. (authors)

  5. ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies. Supplement 1

    International Nuclear Information System (INIS)

    The original ANSL-V cross-section libraries (ORNL-6618) were developed over a period of several years for the physics analysis of the ANS reactor, with little thought toward including the materials commonly needed for shielding applications. Materials commonly used for shielding applications include calcium barium, sulfur, phosphorous, and bismuth. These materials, as well as 6Li, 7Li, and the naturally occurring isotopes of hafnium, have been added to the ANSL-V libraries. The gamma-ray production and gamma-ray interaction cross sections were completely regenerated for the ANSL-V 99n/44g library which did not exist previously. The MALOCS module was used to collapse the 99n/44g coupled library to the 39n/44g broad- group library. COMET was used to renormalize the two-dimensional (2- D) neutron matrix sums to agree with the one-dimensional (1-D) averaged values. The FRESH module was used to adjust the thermal scattering matrices on the 99n/44g and 39n/44g ANSL-V libraries. PERFUME was used to correct the original XLACS Legendre polynomial fits to produce acceptable distributions. The final ANSL-V 99n/44g and 39n/44g cross-section libraries were both checked by running RADE. The AIM module was used to convert the master cross-section libraries from binary coded decimal to binary format (or vice versa)

  6. 78 FR 41436 - Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Science.gov (United States)

    2013-07-10

    ... COMMISSION Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors... Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The NRC seeks public...- Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' This area includes a revised...

  7. Models for transient analyses in advanced test reactors

    International Nuclear Information System (INIS)

    Several strategies are developed worldwide to respond to the world's increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe which employ water as a coolant. Safety provisions and the analyses of severe accidents are key points in the determination of sound designs. In this frame, the SIMMER multiphysics code systems is a very attractive tool as it can simulate transients and phenomena within and beyond the design basis in a tightly coupled way. This thesis is primarily focused upon the extension of the SIMMER multigroup cross-sections processing scheme (based on the Bondarenko method) for a proper heterogeneity treatment in the analyses of water-cooled thermal neutron systems. Since the SIMMER code was originally developed for liquid metal-cooled fast reactors analyses, the effect of heterogeneity had been neglected. As a result, the application of the code to water-cooled systems leads to a significant overestimation of the reactivity feedbacks and in turn to non-conservative results. To treat the heterogeneity, the multigroup cross-sections should be computed by properly taking account of the resonance self-shielding effects and the fine intra-cell flux distribution in space group-wise. In this thesis, significant improvements of the SIMMER cross-section processing scheme are described. A new formulation of the background cross-section, based on the Bell and Wigner correlations, is introduced and pre-calculated reduction factors (Effective Mean Chord Lengths) are used to take proper account of the resonance self-shielding effects of non-fuel isotopes. Moreover, pre-calculated parameters are applied

  8. Models for transient analyses in advanced test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gabrielli, Fabrizio

    2011-02-22

    Several strategies are developed worldwide to respond to the world's increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe which employ water as a coolant. Safety provisions and the analyses of severe accidents are key points in the determination of sound designs. In this frame, the SIMMER multiphysics code systems is a very attractive tool as it can simulate transients and phenomena within and beyond the design basis in a tightly coupled way. This thesis is primarily focused upon the extension of the SIMMER multigroup cross-sections processing scheme (based on the Bondarenko method) for a proper heterogeneity treatment in the analyses of water-cooled thermal neutron systems. Since the SIMMER code was originally developed for liquid metal-cooled fast reactors analyses, the effect of heterogeneity had been neglected. As a result, the application of the code to water-cooled systems leads to a significant overestimation of the reactivity feedbacks and in turn to non-conservative results. To treat the heterogeneity, the multigroup cross-sections should be computed by properly taking account of the resonance self-shielding effects and the fine intra-cell flux distribution in space group-wise. In this thesis, significant improvements of the SIMMER cross-section processing scheme are described. A new formulation of the background cross-section, based on the Bell and Wigner correlations, is introduced and pre-calculated reduction factors (Effective Mean Chord Lengths) are used to take proper account of the resonance self-shielding effects of non-fuel isotopes. Moreover, pre-calculated parameters are

  9. Drop-in capsule testing of plutonium-based fuels in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The most attractive way to dispose of weapons-grade plutonium (WGPu) is to use it as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PuO[sub 2]) mixed with urania (UO[sub 2]). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in- reactor thermal, mechanical, and fission gas release behavior of the prototype fuel will most likely be required in a limited number of test reactor irradiations. The application to license operation with MOX fuel must be amply supported by experimental data. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) is capable of playing a key role in the irradiation, development, and licensing of these new fuel types. The ATR is a 250- MW (thermal) LWR designed to study the effects of intense radiation on reactor fuels and materials. For 25 years, the primary role of the ATR has been to serve in experimental investigations for the development of advanced nuclear fuels. Both large- and small-volume test positions in the ATR could be used for MOX fuel irradiation. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. Furthermore, these data can be obtained more quickly by using ATR instead of testing in a commercial LWR. Our previous work in this area has demonstrated that it is technically feasible to perform MOX fuel testing in the ATR. This report documents our analyses of sealed drop-in capsules containing plutonium-based test specimens placed in various ATR positions

  10. Study of isotopic exchange reactors (1961)

    International Nuclear Information System (INIS)

    A study is made of the general case of the theory of first-order isotopic chemical exchange between a gaseous and a liquid phase in a reactor, starting from fundamental reaction kinetics data, and without making any limiting hypothesis concerning the value of the separation factor. The cases of counter-current reactors and of co-current reactors are considered successively. The general deuterium conservation equation requires the definition of the quotient of the reactor; the performances of this reactor are characterised by its overall efficiency. The idea of the ratio is introduced because it represents a convenient intermediary in the calculations. The search for an additive value for reactors in series leads logically to the defining of an exchange capacity, and a total efficiency, or number of theoretical reactors. This method of expressing the performances of a reactor is more general than the efficiency due to Murphee which only has a physical significance in the particular case of homogeneous liquid reactors. The relationships between these various quantities are established, and the representation due to Mc Cabe and Thiele is generalized. The reactor performances are linked to the first - order reaction kinetics by the transfer number. The relationships are given for a certain number of concrete cases. Finally the application of these calculations is given, together with the approximations necessary in the case where, because of the presence of several components in each phase, the exchange reaction no longer obeys a single kinetic law. (authors)

  11. Extrap conceptual fusion reactor design study

    International Nuclear Information System (INIS)

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

  12. The scaling of economic and performance parameters of DT and advanced fuel fusion reactors

    International Nuclear Information System (INIS)

    In this study, the plasma stability index beta and the fusion power density in the plasma were treated as independent variables to determine how they influenced three economic performance parameters of fusion reactors burning the DT and four advanced fusion fuel cycles. The economic/performance parameters included the total power produced per unit length of reactor; the mass per unit length, and the specific mass in kilograms/kilowatt. The scaling of these parameters with beta and fusion power density was examined for a common set of engineering assumptions on the allowable wall loading limits, the maximum magnetic field existing in the plasma, average blanket mass density, etc. It was found that the power per unit length decreased as the plasma power density and beta increased. This is a consequence of the fact that the first wall is a bottleneck in the energy flow from the plasma to the generating equipment, and the wall power flux will exceed wall loading limits if the plasma radius exceeds a critical value. If one wished to build an engineering test reactor which produced a burning plasma at the lowest possible initial cost, and without regard to whether such a reactor would ultimately produce the cheapest power, then one would minimize the mass per unit length. The mass per unit length decreases with increasing plasma power density and beta, with the DT reaction being the most expensive at a fixed plasma power density (because of its thicker blanket), and the least expensive at a fixed value of beta, at least up to values of beta of 50%. The specific mass, in kg/kw, which is a rough measure of the cost of the power generated by the reactor, shows an opposite trend. It increases with increasing plasma power density and beta. At a given plasma power density and low beta, the DT reaction gives the lowest specific mass, but at a fixed beta above 10%, the advanced fuel cycles have the lowest specific mass

  13. Parametric studies of tandem mirror reactors

    International Nuclear Information System (INIS)

    This report, along with its companion, An Improved Tandem Mirror Reactor, discusses the recent progress and present status of our tandem mirror reactor studies. This report presents the detailed results of parametric studies up to, but not including, the very new ideas involving thermal barriers

  14. The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

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

  15. The advanced test reactor national scientific user facility advancing nuclear technology

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

  16. Reference site selection report for the advanced liquid metal reactor at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    This Reference Site Selection Report was prepared by EG ampersand G, Idaho Inc., for General Electric (GE) to provide information for use by the Department of Energy (DOE) in selecting a Safety Test Site for an Advanced Liquid Metal Reactor. Similar Evaluation studies are planned to be conducted at other potential DOE sites. The Power Reactor Innovative Small Module (PRISM) Concept was developed for ALMR by GE. A ALMR Safety Test is planned to be performed on a DOE site to demonstrate features and meet Nuclear Regulatory Commission Requirements. This study considered possible locations at the Idaho National Engineering Laboratory that met the ALMR Prototype Site Selection Methodology and Criteria. Four sites were identified, after further evaluation one site was eliminated. Each of the remaining three sites satisfied the criteria and was graded. The results were relatively close. Thus concluding that the Idaho National Engineering Laboratory is a suitable location for an Advanced Liquid Metal Reactor Safety Test. 23 refs., 13 figs., 9 tabs

  17. The benefits of an advanced fast reactor fuel cycle for plutonium management

    International Nuclear Information System (INIS)

    The United States has no program to investigate advanced nuclear fuel cycles for the large-scale consumption of plutonium from military and civilian sources. The official U.S. position has been to focus on means to bury spent nuclear fuel from civilian reactors and to achieve the spent fuel standard for excess separated plutonium, which is considered by policy makers to be an urgent international priority. Recently, the National Research Council published a long awaited report on its study of potential separation and transmutation technologies (STATS), which concluded that in the nuclear energy phase-out scenario that they evaluated, transmutation of plutonium and long-lived radioisotopes would not be worth the cost. However, at the American Nuclear Society Annual Meeting in June, 1996, the STATS panelists endorsed further study of partitioning to achieve superior waste forms for burial, and suggested that any further consideration of transmutation should be in the context of energy production, not of waste management. 2048 The U.S. Department of Energy (DOE) has an active program for the short-term disposition of excess fissile material and a 'focus area' for safe, secure stabilization, storage and disposition of plutonium, but has no current programs for fast reactor development. Nevertheless, sufficient data exist to identify the potential advantages of an advanced fast reactor metallic fuel cycle for the long-term management of plutonium. Advantages are discussed

  18. 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)

  19. 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)

  20. Technical Basis for Physical Fidelity of NRC Control Room Training Simulators for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Minsk, Brian S.; Branch, Kristi M.; Bates, Edward K.; Mitchell, Mark R.; Gore, Bryan F.; Faris, Drury K.

    2009-10-09

    The objective of this study is to determine how simulator physical fidelity influences the effectiveness of training the regulatory personnel responsible for examination and oversight of operating personnel and inspection of technical systems at nuclear power reactors. It seeks to contribute to the U.S. Nuclear Regulatory Commission’s (NRC’s) understanding of the physical fidelity requirements of training simulators. The goal of the study is to provide an analytic framework, data, and analyses that inform NRC decisions about the physical fidelity requirements of the simulators it will need to train its staff for assignment at advanced reactors. These staff are expected to come from increasingly diverse educational and experiential backgrounds.

  1. One-dimensional TRAC calculations of main steam line break events for the updated PIUS 600 advanced reactor design

    International Nuclear Information System (INIS)

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. As part of the preapplication and eventual design certification process, advanced reactor applicants are required to submit neutronic and thermal-hydraulic safety analyses over a sufficient range of normal operation, transient conditions, and specified accident sequences. Los Alamos is supporting the US Nuclear Regulatory Commission's preapplication review of the PIUS reactor. A fully one-dimensional model of the PIUS reactor has been developed for the Transient Reactor Analysis Code, TRACPF1/MOD2. Early in 1992, ABB submitted a Supplemental Information Package describing recent design modifications. An important feature of the PIUS Supplement design was the addition of an active scram system that will function for most transient and accident conditions. A one-dimensional Transient Reactor Analysis Code baseline calculation of the PIUS Supplement design were performed for a break in the main steam line at the outlet nozzle of the loop 3 steam generator. Sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions following a main steam line break. The sensitivity study results provide insights into the robustness of the design

  2. Studies to demonstrate the adequacy of testing results of the qualification tests for the actuator of main steam safety relive valves (MSSRV) in an advanced boiling water reactor (ABWR)

    International Nuclear Information System (INIS)

    This paper presents several studies performed to demonstrate that the testing results from the qualification tests for the actuator of the Main Steam Safety Relief Valves (MSSRV; also called SRV in this paper) in GE's Advanced Boiling Water Reactor (ABWR) are in compliance with the qualification guidelines stipulated in the applicable IEEE standards. The safety-related function of the MSSRV is to relieve pressure in order to protect the reactor pressure vessel from over-pressurization condition during normal operation and design basis events. In order to perform this function, the SRV must actuate at a given set pressure while maintaining the pressure and structural integrity of the SRV. The valves are provided with an electro-pneumatic actuator assembly that opens the valve upon receipt of an automatic or manually initiated electric signal to allow depressurization of the reactor pressure vessel (RPV). To assure the SRV can perform its intended safety related functions properly, qualification tests are needed in addition to analysis, to demonstrate that the SRV can withstand the specified environmental, dynamic and seismic design basis conditions without impairing its safety related function throughout their installed life under the design conditions including postulated design basis events such as OBE loads and Faulted (SSE) events. The guidelines used for the test methods, procedures and acceptance criteria for the qualification tests are established in IEEE std 344-1987 and IEEE std 382-1985. In the qualification tests, the specimen consists of the actuator, control valve assembly, limit switches, and limit switch support structure. During the functional, dynamic and seismic tests, the test specimen was mounted on a SRV. Qualification of safety related equipment to meet the guidelines of the IEEE standards is typically a two-step process: 1) environmental aging and 2) design basis events qualification. The purpose of the first step is to put the equipment in an

  3. A comparative study of various advanced fusions

    International Nuclear Information System (INIS)

    For the purpose of comparing the merits and demerits of various advanced fuel cycles, parametric studies of operation conditions are examined. The effects of nuclear elastic collisions and synchrotron radiation are taken into account. It is found that the high-#betta# Catalyzed DD fuel cycle with the transmutation of fusion-produced tritium into helium-3 is most feasible from the point of view of neutron production and tritium handling. The D-D fuel cycles seem to be less attractive compared to the Catalyzed DD. The p-11B and p-6Li fusion plasmas hardly attain the plasma Q value relevant to reactors. (author)

  4. The IRIS consortium: international cooperation in advanced reactor development

    International Nuclear Information System (INIS)

    Besides its many outstanding technical innovations in the design and safety, the most innovative feature of the International Reactor Innovative and Secure (IRIS), is perhaps the international cooperation which carries on its development. IRIS is designed by an international consortium which currently numbers 21 organizations from ten countries across four continents. It includes reactor, fuel and fuel cycle vendors, component manufacturers, laboratories, academia, architect engineers and power producers. The defining organizational characteristics of IRIS is that while Westinghouse has overall lead and responsibility, this lead is of the type of 'primus inter pares' (first among equals) rather than the traditional owner versus suppliers/contractors relationship. All members of the IRIS consortium contribute and expect to have a return, should IRIS be successfully deployed, commensurate to their investment. The nature of such return will be tailored to the type of each organization, because it will of course be of a different nature for say a component manufacturer, university, or architect engineer. One fundamental tenet of the consortium is that all members, regardless of their amount of contribution, have equal access to all information developed within the project. Technical work is thus being coordinated by integrated subgroups and the whole team meets twice a year to perform an overall review of the work, discuss policy and strategy and plan future activities. Personnel from consortium members have performed internships, mostly at Westinghouse locations in Pittsburgh, Pennsylvania, and Windsor, Connecticut, but also at other members, as it has been the case for several graduate students. In fact, more than one hundred students at the various universities have been working on IRIS, most of them conducting graduate theses at the master or doctoral level. The IRIS experience has proved very helpful to the students in successfully landing their employment choice

  5. CARNSORE: Hypothetical reactor accident study

    International Nuclear Information System (INIS)

    Two types of design-basis accident and a series of hypothetical core-melt accidents to a 600 MWe reactor are described and their consequences assessed. The PLUCON 2 model was used to calculate the consequences which are presented in terms of individual and collective doses, as well as early and late health consequences. The site proposed for the nucelar power station is Carnsore Point, County Wexford, south-east Ireland. The release fractions for the accidents described are those given in WASH-1400. The analyses are based on the resident population as given in the 1979 census and on 20 years of data from the meteorological stations at Rosslare Harbour, 8.5 km north of the site. The consequences of one of the hypothetical core-melt accidents are described in detail in a meteorological parametric study. Likewise the consequences of the worst conceivable combination of situations are described. Finally, the release fraction in one accident is varied and the consequences of a proposed, more probable ''Class 9 accident'' are presented. (author)

  6. 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

  7. Analysis of core calculation schemes for advanced water reactors

    International Nuclear Information System (INIS)

    This research thesis addresses the analysis of the core control of sub-moderated water reactors with plutonium fuel and varying spectrum. Firstly, a calculation scheme is defined, based on transport theory for the three existing assembly configurations. It is based on the efficiency analysis of the control cluster and of the flow sheet shape in the assembly. Secondly, studies of the assembly with control cluster and within a theory of diffusion with homogenization or detailed assembly representation are performed by taking the environment into account in order to assess errors. Thirdly, due to the presence of a very efficient absorbent in control clusters, a deeper physical analysis requires the study of the flow gradient existing at the interface between assemblies. A parameter is defined to assess this gradient, and theoretically calculated by using finite elements. Developed software is validated

  8. Measurement of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Sulphur is an important element in some food chains and the release of radioactive sulphur to the environment must be closely controlled if the chemical form is such that it is available or potentially available for entering food chains. The presence of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor warranted a study to assess the quantity and chemical form of the radioactive sulphur in order to estimate the magnitude of the potential environmental hazard which might arise from the release of coolant gas from Civil Advanced Gas-Cooled Reactors. A combination of gas chromatographic and radiochemical analyses revealed carbonyl sulphide to be the only sulphur-35 compound present in the coolant gas of the Windscale Reactor. The concentration of carbonyl sulphide was found to lie in the range 40 to 100 x 10-9 parts by volume and the sulphur-35 specific activity was about 20 mCi per gramme. The analytical techniques are described in detail. The sulphur-35 appears to be derived from the sulphur and chlorine impurities in the graphite. A method for the preparation of carbonyl sulphide labelled with sulphur-35 is described. (author)

  9. Thorium fuel studies for CANDU reactors

    International Nuclear Information System (INIS)

    Applying the once-through Thorium (OTT) cycle in existing and advanced CANDU reactors might be seen as an evolved concept for the sustainable development both from the economic and waste management points of view. Using the Canadian proposed scheme - loading mixed ThO2-SEU CANFLEX bundles in CANDU 6 reactors - simulated at lattice cell level led to promising conclusions on higher burnup, lesser actinide inventory and proliferation resistance. The calculations were performed using the lattice codes WIMS and DRAGON (together with the corresponding nuclear data library based on ENDF/B-VII). (authors)

  10. Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A L; Cetiner, M S; Wilson, Jr, T L

    2012-04-30

    The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary

  11. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    International Nuclear Information System (INIS)

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC

  12. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Seo, K. W

    2006-01-15

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC.

  13. Conceptual Design of Primary Cooling System for an Advanced HANARO Reactor

    International Nuclear Information System (INIS)

    An advanced HANARO Reactor (AHR) is an open-tank-type and generates a 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 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. Through this study, the conceptual design of the primary cooling system has been established including design requirements, performance requirements, design restrictions, system descriptions and system operation to maintain the system function. And preliminary design requirement of the primary cooling system has been established in based on the conceptual design

  14. 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

  15. 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.

  16. Developing a Comprehensive Software Suite for Advanced Reactor Performance and Safety Analysis

    International Nuclear Information System (INIS)

    This paper provides an introduction to the reactor analysis capabilities of the nuclear power reactor simulation tools that are being developed as part of the US Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Toolkit. The NEAMS Toolkit is an integrated suite of multiphysics simulation tools that leverage high performance computing to reduce uncertainty in the prediction of the performance and safety of advanced reactor and fuel designs. The toolkit effort is composed of two major components, the fuels product line, which provides tools for fuel performance analysis, and the reactor product line, which provides tools for reactor performance and safety analysis. This paper presents an overview of the NEAMS reactor product line development effort. (author)

  17. Development of the Advanced CANDU Reactor control centre

    International Nuclear Information System (INIS)

    The next generation CANDU control centre is being designed for the Advanced CANDU Reactor (ACR) station. The design is based upon the recent Qinshan control room with further upgrades to meet customer needs with respect to high capacity factor with low Operation, Maintenance and Administration (OM and A) costs. This evolutionary design includes the long proven functionality at several existing CANDU control centres such as the 4-unit station at Darlington, with advanced features made possible by new control and display technology. Additionally, ACR control centres address characteristics resulting from Human Factors Engineering (HFE) analysis of control centre operations in order to further enhance personnel awareness of system and plant status. Statistics show that up to 70% of plant significant events, which have caused plant outages, have a root cause attributable to the human from such sources as complex interfaces, procedures, maintenance and management practices. Consequently, special attention is made for the application of HFE throughout the ACR design process. The design process follows a systematic analytical approach to define operations staff information and information presentation requirements. The resultant human-system interfaces (HSI) such as those for monitoring, annunciation and control information are then verified and validated against the system design requirements to provide a high confidence level that adequate and correct information is being provided in a timely manner to support the necessary operational tasks. The ACR control centre provides plant staff with an improved operability capability due to the combination of systematic design and enhanced operating features. Significant design processes (i.e. development) or design features which contribute to this improved operability, include: Design Process: Project HFE Program Plan - intent, scope, timeliness and interfacing; HFE aspects of design process - procedures and instructions

  18. Development of the advanced CANDU reactor control centre

    International Nuclear Information System (INIS)

    The next generation CANDU control centre is being designed for the Advanced CANDU Reactor (ACR) station. The design is based upon the recent Qinshan control room with further upgrades to meet customer needs with respect to high capacity factor with low Operation, Maintenance and Administration (OM and A) costs. This evolutionary design includes the long proven functionality at several existing CANDU control centres such as the 4-unit station at Darlington, with advanced features made possible by new control and display technology. Additionally, ACR control centres address characteristics resulting from Human Factors Engineering (HFE) analysis of control centre operations in order to further enhance personnel awareness of system and plant status. Statistics show that up to 70% of plant significant events