Science.gov

Sample records for plant reactors designed

  1. Generic small modular reactor plant design.

    SciTech Connect

    Lewis, Tom Goslee,; Cipiti, Benjamin B.; Jordan, Sabina Erteza; Baum, Gregory A.

    2012-12-01

    This report gives an overview of expected design characteristics, concepts, and procedures for small modular reactors. The purpose of this report is to provide those who are interested in reducing the cost and improving the safety of advanced nuclear power plants with a generic design that possesses enough detail in a non-sensitive manner to give merit to their conclusions. The report is focused on light water reactor technology, but does add details on what could be different in a more advanced design (see Appendix). Numerous reactor and facility concepts were used for inspiration (documented in the bibliography). The final design described here is conceptual and does not reflect any proposed concept or sub-systems, thus any details given here are only relevant within this report. This report does not include any design or engineering calculations.

  2. Design considerations for an inertial confinement fusion reactor power plant

    SciTech Connect

    Massey, J.V.; Simpson, J.E.

    1981-08-10

    To further define the engineering and economic concerns for inertial confinement fusion reactors (ICR's), a conceptual design study was performed by Bechtel Group Incorporated under the direction of Lawrence Livermore National Laboratory (LLNL). The study examined alternatives to the LLNL HYLIFE concept and expanded the previous balance of plant design to incorporate information from recent liquid metal cooled fast breeder reactor (LMFBR) power plant studies. The majority of the effort was to incorporate present laser and target physics models into a reactor design with a low coolant flowrate and a high driver repetition rate. An example of such a design is the LLNL JADE concept. In addition to producing a power plant design for LLNL using the JADE example, Bechtel has also examined the applicability of the EAGLE (Energy Absorbing Gas Lithium Ejector) concept.

  3. Fast reactor power plant design having heat pipe heat exchanger

    DOEpatents

    Huebotter, P.R.; McLennan, G.A.

    1984-08-30

    The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

  4. Fast reactor power plant design having heat pipe heat exchanger

    DOEpatents

    Huebotter, Paul R.; McLennan, George A.

    1985-01-01

    The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

  5. Overall plant design specification Modular High Temperature Gas-cooled Reactor. Revision 9

    SciTech Connect

    1990-05-01

    Revision 9 of the ``Overall Plant Design Specification Modular High Temperature Gas-Cooled Reactor,`` DOE-HTGR-86004 (OPDS) has been completed and is hereby distributed for use by the HTGR Program team members. This document, Revision 9 of the ``Overall Plant Design Specification`` (OPDS) reflects those changes in the MHTGR design requirements and configuration resulting form approved Design Change Proposals DCP BNI-003 and DCP BNI-004, involving the Nuclear Island Cooling and Spent Fuel Cooling Systems respectively.

  6. Helium circulator design considerations for modular high temperature gas-cooled reactor plant

    SciTech Connect

    McDonald, C.F.; Nichols, M.K.

    1986-12-01

    Efforts are in progress to develop a standard modular high temperature gas-cooled reactor (MHTGR) plant that is amenable to design certification and serial production. The MHTGR reference design, based on a steam cycle power conversion system, utilizes a 350 MW(t) annular reactor core with prismatic fuel elements. Flexibility in power rating is afforded by utilizing a multiplicity of the standard module. The circulator, which is an electric motor-driven helium compressor, is a key component in the primary system of the nuclear plant, since it facilitates thermal energy transfer from the reactor core to the steam generator; and, hence, to the external turbo-generator set. This paper highlights the helium circulator design considerations for the reference MHTGR plant and includes a discussion on the major features of the turbomachine concept, operational characteristics, and the technology base that exists in the US.

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

    SciTech Connect

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

    1982-05-01

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

  8. Slurry reactor design studies

    SciTech Connect

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. ); Akgerman, A. ); Smith, J.M. )

    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.

  9. Maintenance Cycle Extension in the IRIS Advanced Light Water Reactor Plant Design

    SciTech Connect

    Galvin, Mark R.; Todreas, Neil E.; Conway, Larry E.

    2003-09-15

    New nuclear power generation in the United States will be realized only if the economic performance can be made competitive with other methods of electrical power generation. The economic performance of a nuclear power plant can be significantly improved by increasing the time spent on-line generating electricity relative to the 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 that can be used to resolve, in the design phase, maintenance-related operating cycle length barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the International Reactor, Innovative and Secure (IRIS) design. IRIS is an advanced light water nuclear power plant that is being designed to maximize this on-line generating time by increasing the operating cycle length. This is consequently a maintenance strategy paper using the IRIS plant as the example.Potential IRIS operating cycle length maintenance-related barriers, determined by modification of an earlier operating pressurized water reactor (PWR) plant cycle length analysis to account for differences between the design of IRIS and this operating PWR, are presented. The proposed methodology to resolve these maintenance-related barriers by the design process is described. The results of applying the methodology to two potential IRIS cycle length barriers, relief valve testing and emergency heat removal system testing, are presented.

  10. NEUTRONIC REACTOR POWER PLANT

    DOEpatents

    Metcalf, H.E.

    1962-12-25

    This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

  11. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    NASA Astrophysics Data System (ADS)

    Ilham, Muhammad; Su’ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  12. Westinghouse Small Modular Reactor balance of plant and supporting systems design

    SciTech Connect

    Memmott, M. J.; Stansbury, C.; Taylor, C.

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

  13. Modular pebble-bed reactor reforming plant design for process heat

    SciTech Connect

    Lutz, D.E.; Cowan, C.L.; Davis, C.R.; El Sheikh, K.A.; Hui, M.M.; Lipps, A.J.; Wu, T.

    1982-09-01

    This report describes a preliminary design study of a Modular Pebble-Bed Reactor System Reforming (MPB-R) Plant. The system uses one pressure vessel for the reactor and a second pressure vessel for the components, i.e., reformer, steam generator and coolant circulator. The two vessels are connected by coaxial pipes in an arrangement known as the side-by-side (SBS). The goal of the study is to gain an understanding of this particular system and to identify any technical issues that must be resolved for its application to a modular reformer plant. The basic conditions for the MPB-R were selected in common with those of the current study of the MRS-R in-line prismatic fuel concept, specifically, the module core power of 250 MWt, average core power density of 4.1 w/cc, low enriched uranium (LEU) fuel with a /sup 235/U content of 20% homogeneously mixed with thorium, and a target burnup of 80,000 MWD/MT. Study results include the pebble-bed core neutronics and thermal-hydraulic calculations. Core characteristics for both the once-through-then-out (OTTO) and recirculation of fuel sphere refueling schemes were developed. The plant heat balance was calculated with 55% of core power allotted to the reformer.

  14. Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

    NASA Astrophysics Data System (ADS)

    Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

    2016-01-01

    Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

  15. Advanced Light Water Reactor Plants System 80+{trademark} Design Certification Program. Annual progress report, October 1, 1992--September 30, 1993

    SciTech Connect

    Not Available

    1993-12-31

    The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1993 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW{sub t} (1350 MWe) Pressurized Water Reactor (PWR). The design consists of an essentially complete plant. It is based on evolutionary improvements to the Standardized System 80 nuclear steam supply system in operation at Palo Verde Units 1, 2, and 3, and the Duke Power Company P-81 balance-of-plant (BOP) that was designed and partially constructed at the Cherokee plant site. The System 80/P-81 original design has been substantially enhanced to increase conformance with the EPRI ALWR Utility Requirements Document (URD). Some design enhancements incorporated in the System 80+ design are included in the four units currently under construction in the Republic of Korea. These units form the basis of the Korean standardization program. The full System 80+ standard design has been offered to the Republic of China, in response to their recent bid specification. The ABB-CE Standard Safety Analysis Report (CESSAR-DC) was submitted to the NRC and a Draft Safety Evaluation Report was issued by the NRC in October 1992. CESSAR-DC contains the technical basis for compliance with the EPRI URD for simplified emergency planning. The Nuclear Steam Supply System (NSSS) is the standard ABB-Combustion Engineering two-loop arrangement with two steam generators, two hot legs and four cold legs each with a reactor coolant pump. The System 80+ standard plant includes a sperical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual containment.

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

    SciTech Connect

    Not Available

    1983-01-01

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

  17. Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2009-05-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep 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 graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any

  18. Advanced light water reactor plants System 80+{trademark} design certification program. Annual progress report, October 1, 1995--September 30, 1996

    SciTech Connect

    1996-12-31

    The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1996 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2 and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems.

  19. Analysis and selection of high pressure heaters design for a new generation of NPP with BN-1200 reactor plant

    NASA Astrophysics Data System (ADS)

    Yurchenko, A. Yu.; Sukhorukov, Yu. G.; Trifonov, N. N.; Grigor'eva, E. B.; Esin, S. B.; Svyatkin, F. A.; Nikolaenkova, E. K.; Prikhod'ko, P. Yu.; Nazarov, V. V.

    2016-09-01

    In the development of advanced high-power steam-turbine plants (STP), special attention is placed on the design of reliable and economical high-pressure heater (HPH) capable to maintain the specified thermal hydraulic performance during the entire service life. Comparative analysis of the known designs of HPH, such as the spiral-collector HPH, the collector-coiled HPH, the collector-platen HPH, modular HPH, and the chamber HPH, was carried out. The advantages and disadvantages of each design were pointed. For better comparison, the heaters are separated into two groups—horizontal and vertical ones. The weight and dimension characteristics, the materials and features of the basic elements, and operating features of variety HPH are presented. At operating the spiral-collector HPH used in the majority of regenerative schemes of high-pressure STP of thermal and nuclear power plants, the disadvantages reducing the economy and reliability of their operation were revealed. The recommendations directed to the reliability growth of HPH, the decrease of subcooling the feed water, the increase of compactness are stated. Some of these were developed by the specialists of OAO NPO TsKTI and are successfully implemented on the thermal power plants and nuclear power plants. Technical solutions to reduce the cost of regeneration system and the weight of chamber HPH, reduce the thickness of the tube plate of HPH, and reliability assurance of the cooler of steam and condensate built in the HPH casing under all operating conditions were proposed. Three types of feed water chambers for vertical and horizontal chamber HPH are considered in detail, the constructive solutions that have been implemented in HPH of the regeneration system of turbines of 1000 and 1200 MW capacity with water-moderated water-cooled power reactor (WMWCPR) are described. The optimal design of HPH for the regeneration system of high-pressure turbine plant with BN-1200 reactor was selected.

  20. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    NASA Technical Reports Server (NTRS)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  1. Safety design of prototype fast breeder reactor

    SciTech Connect

    Bhoje, S.B.; Chetal, S.C.; Singh, Om Pal

    2004-07-01

    The basic design and safety design of Prototype Fast Breeder Reactor (PFBR) is presented. Design aspects covered include safety classification, seismic categorization, design basis conditions, design safety limits, core physics, core monitoring, shutdown system, decay heat removal system, protection against sodium leaks and tube leaks in steam generator, plant layout, radiation protection, event analysis, beyond design basis accidents, integrity of primary containment, reactor containment building and design pressure resulting from core disruptive accident. The measures provided in the design represent a robust case of the safety of the reactor. (authors)

  2. Thermionic Reactor Design Studies

    SciTech Connect

    Schock, Alfred

    1994-08-01

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  3. Analysis of UF6 breeder reactor power plants

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.

    1976-01-01

    Gaseous UF6 fueled breeder reactor design and technical applications of such concepts are summarized. Special attention was given to application in nuclear power plants and to reactor efficiency and safety factors.

  4. Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

    NASA Technical Reports Server (NTRS)

    Larson, V. R.; Gunn, S. V.; Lee, J. C.

    1975-01-01

    The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

  5. Thermionic Reactor Design Studies

    SciTech Connect

    Schock, Alfred

    1994-06-01

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full

  6. NUCLEAR REACTOR CORE DESIGN

    DOEpatents

    Mahlmeister, J.E.; Peck, W.S.; Haberer, W.V.; Williams, A.C.

    1960-03-22

    An improved core design for a sodium-cooled, graphitemoderated nuclear reactor is described. The improved reactor core comprises a number of blocks of moderator material, each block being in the shape of a regular prism. A number of channels, extending the length of each block, are disposed around the periphery. When several blocks are placed in contact to form the reactor core, the channels in adjacent blocks correspond with each other to form closed conduits extending the length of the core. Fuel element clusters are disposed in these closed conduits, and liquid coolant is forced through the annulus between the fuel cluster and the inner surface of the conduit. In a preferred embodiment of the invention, the moderator blocks are in the form of hexagonal prisms with longitudinal channels cut into the corners of the hexagon. The main advantage of an "edge-loaded" moderator block is that fewer thermal neutrons are absorbed by the moderator cladding, as compared with a conventional centrally loaded moderator block.

  7. MLTAP. Modular Helium Reactor Plant Transient Thermal-Hydraulic Analysis

    SciTech Connect

    Chan, T.W.; Openshaw, F.L.

    1992-11-06

    MLTAP is an integrated system transient analysis code for modular helium reactor (MHR) plants with superheated steam for Rankine power cycle and/or process heat applications. It is used for normal operational transient analyses as well as design basis/accident condition analyses with forced convection reactor cooling. MLTAP calculates the time-dependent temperatures, pressures, and flow rates for helium primary coolant and steam/water secondary coolant; reactor system and steam system structural temperatures; reactor neutronic behavior; pump, compressor, and steam turbine performance; reactivity control and other plant control systems responses; reactor and plant protection systems responses.

  8. 75 FR 61227 - Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Future Plant Designs...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-04

    ..., General Electric--Hitachi Nuclear Energy (GEH), and their contractors, pursuant to 5 U.S.C. 552b(c)(4... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Future Plant...

  9. Turning points in reactor design

    SciTech Connect

    Beckjord, E.S.

    1995-09-01

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

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

    SciTech Connect

    Not Available

    1994-08-01

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

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

    SciTech Connect

    Not Available

    1994-08-01

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

  12. Contribution of Clinch River Breeder Reactor plant design and development to the LMFBR fuel cycle

    SciTech Connect

    Riley, D.R.; Dickson, P.W.

    1981-01-01

    This paper describes how the CRBRP development and CRBRP focus of the LMFBR base technology program have led to advances in the state of the art in physics, thermal-hydraulics, structural analysis, core restraint, seismic analysis, and analysis of hypothetical core-disruptive accident energetics, all of which have been incorporated through disciplined engineering into the final CRBRP design. The total development in the US of fuels and materials, the analytical advances made on CRBRP design, and the incorporation of the latest experimental results into that design have put the US technology in general and the CRBRP design in particular at the forefront of technology. This has placed the US in a position to develop the most favorable LMFBR fuel cycle.

  13. Plant maintenance and advanced reactors, 2007

    SciTech Connect

    Agnihotri, Newal

    2007-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: A new day for energy in America; Committed to success more than ever, by Andy White, GE--Hitachi Nuclear Energy; Competitive technology for decades, by Steve Tritch, Westinghouse Electric Company; Pioneers of positive community relationship, by Exelon Nuclear; A robust design for 60-years, by Ray Ganthner, Areva; Aiming at no evacuation plants, by Kumiaki Moriya, Hitachi-GE Nuclear Energy, Ltd.; and, Desalination and hydrogen economy, by Dr. I. Khamis, International Atomic Energy Agency. Industry innovation articles in this issue are: Reactor vessel closure head project, by Jeff LeClair, Prairie Island Nuclear Generating Plant; and Submersible remote-operated vehicle, by Michael S. Rose, Entergy's Fitzpatrick Nuclear Station.

  14. Modular stellarator reactor: a fusion power plant

    SciTech Connect

    Miller, R.L.; Bathke, C.G.; Krakowski, R.A.; Heck, F.M.; Green, L.; Karbowski, J.S.; Murphy, J.H.; Tupper, R.B.; DeLuca, R.A.; Moazed, A.

    1983-07-01

    A comparative analysis of the modular stellarator and the torsatron concepts is made based upon a steady-state ignited, DT-fueled, reactor embodiment of each concept for use as a central electric-power station. Parametric tradeoff calculations lead to the selection of four design points for an approx. 4-GWt plant based upon Alcator transport scaling in l = 2 systems of moderate aspect ratio. The four design points represent high-aspect ratio. The four design points represent high-(0.08) and low-(0.04) beta versions of the modular stellarator and torsatron concepts. The physics basis of each design point is described together with supporting engineering and economic analyses. The primary intent of this study is the elucidation of key physics and engineering tradeoffs, constraints, and uncertainties with respect to the ultimate power reactor embodiment.

  15. FASTER Test Reactor Preconceptual Design Report

    SciTech Connect

    Grandy, C.; Belch, H.; Brunett, A. J.; Heidet, F.; Hill, R.; Hoffman, E.; Jin, E.; Mohamed, W.; Moisseytsev, A.; Passerini, S.; Sienicki, J.; Sumner, T.; Vilim, R.; Hayes, S.

    2016-03-31

    The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  16. FASTER test reactor preconceptual design report summary

    SciTech Connect

    Grandy, C.; Belch, H.; Brunett, A.; Heidet, F.; Hill, R.; Hoffman, E.; Jin, E.; Mohamed, W.; Moisseytsev, A.; Passerini, S.; Sienicki, J.; Sumner, T.; Vilim, R.; Hayes, Steven

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  17. Design analysis of the molten core confinement within the reactor vessel in the case of severe accidents at nuclear power plants equipped with a reactor of the VVER type

    NASA Astrophysics Data System (ADS)

    Zvonaryov, Yu. A.; Budaev, M. A.; Volchek, A. M.; Gorbaev, V. A.; Zagryazkin, V. N.; Kiselyov, N. P.; Kobzar', V. L.; Konobeev, A. V.; Tsurikov, D. F.

    2013-12-01

    The present paper reports the results of the preliminary design estimate of the behavior of the core melt in vessels of reactors of the VVER-600 and VVER-1300 types (a standard optimized and informative nuclear power unit based on VVER technology—VVER TOI) in the case of beyond-design-basis severe accidents. The basic processes determining the state of the core melt in the reactor vessel are analyzed. The concept of molten core confinement within the vessel based on the idea of outside cooling is discussed. Basic assumptions and models, as well as the results of calculation of the interaction between molten materials of the core and the wall of the reactor vessel performed by means of the SOCRAT severe accident code, are presented and discussed. On the basis of the data obtained, the requirements on the operation of the safety systems are determined, upon the fulfillment of which there will appear potential prerequisites for implementing the concept of the confinement of the core melt within the reactor in cases of severe accidents at nuclear power plants equipped with VVER reactors.

  18. Lead-Cooled Fast Reactor (LFR) Design: Safety, Neutronics, Thermal Hydraulics, Structural Mechanics, Fuel, Core, and Plant Design

    SciTech Connect

    Smith, C

    2010-02-22

    The idea of developing fast spectrum reactors with molten lead (or lead alloy) as a coolant is not a new one. Although initially considered in the West in the 1950s, such technology was not pursued to completion because of anticipated difficulties associated with the corrosive nature of these coolant materials. However, in the Soviet Union, such technology was actively pursued during the same time frame (1950s through the 1980s) for the specialized role of submarine propulsion. More recently, there has been a renewal of interest in the West for such technology, both for critical systems as well as for Accelerator Driven Subcritical (ADS) systems. Meanwhile, interest in the former Soviet Union, primarily Russia, has remained strong and has expanded well beyond the original limited mission of submarine propulsion. This section reviews the past and current status of LFR development.

  19. 75 FR 5632 - Office of New Reactors; Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-03

    ...The NRC staff is soliciting public comment on its proposed Interim Staff Guidance (ISG) DC/COL-ISG-021 titled ``Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using a Gas Turbine Driven Standby Emergency Alternating Current Power System,'' (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092640035). This ISG provides new guidance information for......

  20. Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H. Volume 1, Final report

    SciTech Connect

    Waganer, L.M.; Driemeyer, D.E.; Lee, V.D.

    1992-03-01

    This report contains a review of design studies for inertial confinement reactors. The first of three volumes briefly discusses the following: Introduction; Key objectives, requirements, and assumptions; Systems modeling and trade studies; Prometheus-L reactor plant design overview; Prometheus-H reactor plant design overview; Key technical issues and R&D requirements; Comparison of IFE designs; and study conclusions.

  1. Simplifying microbial electrosynthesis reactor design

    PubMed Central

    Giddings, Cloelle G. S.; Nevin, Kelly P.; Woodward, Trevor; Lovley, Derek R.; Butler, Caitlyn S.

    2015-01-01

    Microbial electrosynthesis, an artificial form of photosynthesis, can efficiently convert carbon dioxide into organic commodities; however, this process has only previously been demonstrated in reactors that have features likely to be a barrier to scale-up. Therefore, the possibility of simplifying reactor design by both eliminating potentiostatic control of the cathode and removing the membrane separating the anode and cathode was investigated with biofilms of Sporomusa ovata. S. ovata reduces carbon dioxide to acetate and acts as the microbial catalyst for plain graphite stick cathodes as the electron donor. In traditional ‘H-cell’ reactors, where the anode and cathode chambers were separated with a proton-selective membrane, the rates and columbic efficiencies of microbial electrosynthesis remained high when electron delivery at the cathode was powered with a direct current power source rather than with a potentiostat-poised cathode utilized in previous studies. A membrane-less reactor with a direct-current power source with the cathode and anode positioned to avoid oxygen exposure at the cathode, retained high rates of acetate production as well as high columbic and energetic efficiencies. The finding that microbial electrosynthesis is feasible without a membrane separating the anode from the cathode, coupled with a direct current power source supplying the energy for electron delivery, is expected to greatly simplify future reactor design and lower construction costs. PMID:26029199

  2. 76 FR 17160 - Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-28

    ...The NRC staff is issuing its Final Interim Staff Guidance (ISG) DC/COL-ISG-021 titled ``Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using a Gas Turbine Driven Standby Emergency Alternating Current Power System,'' Agencywide Documents Access and Management System (ADAMS) Accession No. ML102510119 for DC/ COL-ISG-021 and ADAMS Accession No. ML102510164 for Attachment 1 to......

  3. Design and startup of a membrane-biological-reactor system at a Ford-engine plant for treating oily wastewater.

    PubMed

    Kim, B R; Anderson, J E; Mueller, S A; Gaines, W A; Szafranski, M J; Bremmer, A L; Yarema, G J; Guciardo, C D; Linden, S; Doherty, T E

    2006-04-01

    A wastewater-treatment facility at Ford (Dearborn, Michigan) was recently upgraded from chemical de-emulsification to ultrafiltration (UF) followed by a membrane-biological reactor (MBR). This paper describes the design, startup, and initial operational performance of the facility. Primary findings are as follows: (1) the MBR proved resilient; (2) the MBR removed approximately 90% of chemical-oxygen demand (COD) after primary UF; (3) the removal of total Kjeldahl nitrogen by MBR appeared to be more sensitive to operating conditions than COD removal; (4) nitrification and denitrification were established in one month; (5) the MBR removed oil and grease and phenolics to below detection levels consistently, in contrast to widely fluctuating concentrations in the past; (6) permeate fluxes of the primary and MBR UF were adversely affected by inadvertent use of a silicone-based defoamer; and (7) zinc concentrations in the effluent increased, which might have been a result of ethylenediaminetetraacetic acid used in membrane washing solutions and/or might have been within typical concentration ranges.

  4. ALMR plant design and performance

    SciTech Connect

    Kwant, W.; Boardman, C.E.; Dayal, Y.; Magee, P.M. )

    1992-01-01

    The advanced liquid-metal reactor (ALMR) plant, sponsored by the US Department of Energy and being developed by a General Electric Company lead industrial team, features simple and reliable safety systems, seismic isolation, passive decay heat removal, passive reactivity control, and substantial margins to structural and fuel damage limits during potential accident situations. These features will result in significant gains for public safety and protection of the owner's investment. Standardized modular construction and extensive factory fabrication will result in a plant design that is economically competitive. The reference commercial ALMR plant utilizes nine reactor modules arranged in three identical 480-MW(electric) power blocks for an overall plant net electrical rating of 1440 MW(electric). Each power block features three identical reactor modules, each with its own steam generator, that jointly supply power to a single turbine generator.

  5. Design of pilot-scale solar photocatalytic reactor for the generation of hydrogen from alkaline sulfide wastewater of sewage treatment plant.

    PubMed

    Priya, R; Kanmani, S

    2013-01-01

    Experiments were conducted for photocatalytic generation of renewable fuel hydrogen from sulphide wastewater from the sewage treatment plant. In this study, pilot-scale solar photocatalytic reactor was designed for treating 1 m3 of sulphide wastewater and also for the simultaneous generation of hydrogen. Bench-scale studies were conducted both in the batch recycle and continuous modes under solar irradiation at similar experimental conditions. The maximum of 89.7% conversion was achieved in the continuous mode. The length of the pilot-scale solar photocatalytic reactor was arrived using the design parameters such as volumetric flow rate (Q) (11 x 10(-2) m3/s), inlet concentration of sulphide ion (C(in)) (28 mol/m3), conversion (89.7%) and average mass flow destruction rate (3.488 x 10(-6) mol/m2 s). The treatment cost of the process was estimated to be 6 US$/m3. This process would be suitable for India like sub-tropical country where sunlight is abundantly available throughout the year.

  6. Core damage frequency (reactor design) perspectives based on IPE results

    SciTech Connect

    Camp, A.L.; Dingman, S.E.; Forester, J.A.

    1996-12-31

    This paper provides perspectives gained from reviewing 75 Individual Plant Examination (IPE) submittals covering 108 nuclear power plant units. Variability both within and among reactor types is examined to provide perspectives regarding plant-specific design and operational features, and C, modeling assumptions that play a significant role in the estimates of core damage frequencies in the IPEs. Human actions found to be important in boiling water reactors (BWRs) and in pressurized water reactors (PWRs) are presented and the events most frequently found important are discussed.

  7. Russian RBMK reactor design information

    SciTech Connect

    Not Available

    1993-11-01

    This document concerns the systems, design, and operations of the graphite-moderated, boiling, water-cooled, channel-type (RBMK) reactors located in the former Soviet Union (FSU). The Russian Academy of Sciences Nuclear Safety Institute (NSI) in Moscow, Russia, researched specific technical questions that were formulated by the Pacific Northwest Laboratory (PNL) and provided detailed technical answers to those questions. The Russian response was prepared in English by NSI in a question-and-answer format. This report presents the results of that technical exchange in the context they were received from the NSI organization. Pacific Northwest Laboratory is generating this document to support the US Department of Energy (DOE) community in responding to requests from FSU states, which are seeking Western technological and financial assistance to improve the safety systems of the Russian-designed reactors. This report expands upon information that was previously available to the United States through bilateral information exchanges, international nuclear society meetings, International Atomic Energy Agency (IAEA) reactor safety programs, and Research and Development Institute of Power Engineering (RDIPE) reports. The response to the PNL questions have not been edited or reviewed for technical consistency or accuracy by PNL staff or other US organizations, but are provided for use by the DOE community in the form they were received.

  8. NRC policy on future reactor designs

    SciTech Connect

    1985-07-01

    On April 13, 1983, the US Nuclear Regulatory Commission issued for public comment a ''Proposed Commission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regulation'' (48 FR 16014). This report presents and discusses the Commission's final version of that policy statement now entitled, ''Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants.'' It provides an overview of comments received from the public and the Advisory Committee on Reactor Safeguards and the staff response to these. In addition to the Policy Statement, the report discusses how the policies of this statement relate to other NRC programs including the Severe Accident Research Program; the implementation of safety measures resulting from lessons learned in the accident at Three Mile Island; safety goal development; the resolution of Unresolved Safety Issues and other Generic Safety Issues; and possible revisions of rules or regulatory requirements resulting from the Severe Accident Source Term Program. Also discussed are the main features of a generic decision strategy for resolving Regulatory Questions and Technical Issues relating to severe accidents; the development and regulatory use of new safety information; the treatment of uncertainty in severe accident decision making; and the development and implementation of a Systems Reliability Program for both existing and future plants to ensure that the realized level of safety is commensurate with the safety analyses used in regulatory decisions.

  9. Development of high-strength concrete mix designs in support of the prestressed concrete reactor vessel design for a HTGR steam cycle/cogeneration plant

    SciTech Connect

    Naus, D.J.; Oland, C.B.

    1985-01-01

    Design optimization studies indicate that a significant reduction in the size of the PCRV for a 2240 MW(t) HTGR plant can be effected through utilization of high-strength concrete in conjunction with large capacity prestressing systems. A three-phase test program to develop and evaluate high-strength concretes (>63.4 MPa) is described. Results obtained under Phase I of the investigation related to materials selection-evaluation and mix design development are presented. 3 refs., 4 figs.

  10. Prism sodium-cooled reactor design and performance

    SciTech Connect

    Kwant, W.; Magee, P.M.; Patel, M.R. )

    1989-01-01

    The Power Reactor Inherently Safe Module (PRISM) program is being conducted at General Electric (GE) under U.S. Department of Energy sponsorship to develop a conceptual design for an advanced sodium-cooled liquid-metal reactor plant. The PRISM design emphasizes inherent safety, modular construction, and factory fabrication. A PRISM power plant includes a number of reactor modules, which will be fabricated in a factory and shipped by whatever combination of barge, rail, and road transport that is most economical for a particular site. The target commercial PRISM plant utilizes nine reactor modules arranged in three identical 465-MW(electric) power blocks for an overall plant net electrical rating of 1395 MW(electric). Each power block has three identical reactor modules, each with its own steam generator, that jointly supply saturated steam to a single turbine generator. The PRISM's features of fewer and simpler safety systems, seismic isolation, passive decay heat removal, inherent reactivity control, and generous margins from structural and fuel damage limits during potential accident situations will result in significant gains in public safety and protection of the owner's investment. The use of standardized modular construction and extensive factory fabrication is resulting in a plant design that is economically competitive against projected coal plants and other nuclear design approaches.

  11. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    SciTech Connect

    Ruger, C.J.; Higgins, J.C.

    1993-11-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970`s and early 1980`s raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants.

  12. Initiating Events for Multi-Reactor Plant Sites

    SciTech Connect

    Muhlheim, Michael David; Flanagan, George F.; Poore, III, Willis P.

    2014-09-01

    Inherent in the design of modular reactors is the increased likelihood of events that initiate at a single reactor affecting another reactor. Because of the increased level of interactions between reactors, it is apparent that the Probabilistic Risk Assessments (PRAs) for modular reactor designs need to specifically address the increased interactions and dependencies.

  13. Zirconium Hydride Space Power Reactor design.

    NASA Technical Reports Server (NTRS)

    Asquith, J. G.; Mason, D. G.; Stamp, S.

    1972-01-01

    The Zirconium Hydride Space Power Reactor being designed and fabricated at Atomics International is intended for a wide range of potential applications. Throughout the program a series of reactor designs have been evaluated to establish the unique requirements imposed by coupling with various power conversion systems and for specific applications. Current design and development emphasis is upon a 100 kilowatt thermal reactor for application in a 5 kwe thermoelectric space power generating system, which is scheduled to be fabricated and ground tested in the mid 70s. The reactor design considerations reviewed in this paper will be discussed in the context of this 100 kwt reactor and a 300 kwt reactor previously designed for larger power demand applications.

  14. Current status of high conversion pressurized water reactor design studies

    SciTech Connect

    Umeoka, T.; Kono, T.; Toyoda, Y.; Ogino, M.; Iwai, S.; Hishida, H.

    1988-01-01

    Preliminary design studies on high conversion pressurized water reactors (HCPWRs) have been completed, and plant design studies are currently being performed to improve the feasibility of HCPWRs. The present status of the feasibility studies is covered, and the related validation tests to be conducted in the coming years are reviewed.

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

    SciTech Connect

    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.

  16. Overall plant concept for a tank-type fast reactor

    SciTech Connect

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

    1984-01-01

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

  17. Design of virtual SCADA simulation system for pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  18. Design of virtual SCADA simulation system for pressurized water reactor

    SciTech Connect

    Wijaksono, Umar Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-08

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  19. Reactor design for nuclear electric propulsion

    NASA Technical Reports Server (NTRS)

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

    1979-01-01

    The paper analyzes the consequences of heat pipe failures, that resulted in modifications to the basic design of a heat-pipe cooled, fast spectrum nuclear reactor and led to consideration of an entirely different core design. The new design features an integral laminated core configuration consisting of alternating layers of UO2 and molybdenum sheets that span the diameter of the core. Design characteristics are presented and compared for two reactors. A conceptual design for a heat exchanger between the core and the thermionic converter assembly is described. This heat exchanger would provide design and fabrication decoupling of these two assemblies.

  20. Dual-phase reactor plant with partitioned isolation condenser

    DOEpatents

    Hui, Marvin M.

    1992-01-01

    A nuclear energy plant housing a boiling-water reactor utilizes an isolation condenser in which a single chamber is partitioned into a distributor plenum and a collector plenum. Steam accumulates in the distributor plenum and is conveyed to the collector plenum through an annular manifold that includes tubes extending through a condenser pool. The tubes provide for a transfer of heat from the steam, forming a condensate. The chamber has a disk-shaped base, a cylindrical sidewall, and a semispherical top. This geometry results in a compact design that exhibits significant performance and cost advantages over prior designs.

  1. Basis for NGNP Reactor Design Down-Selection

    SciTech Connect

    L.E. Demick

    2010-08-01

    The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

  2. Basis for NGNP Reactor Design Down-Selection

    SciTech Connect

    L.E. Demick

    2011-11-01

    The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

  3. Mirror Advanced Reactor Study interim design report

    SciTech Connect

    Not Available

    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.

  4. Design options for a bunsen reactor.

    SciTech Connect

    Moore, Robert Charles

    2013-10-01

    This work is being performed for Matt Channon Consulting as part of the Sandia National Laboratories New Mexico Small Business Assistance Program (NMSBA). Matt Channon Consulting has requested Sandia's assistance in the design of a chemical Bunsen reactor for the reaction of SO2, I2 and H2O to produce H2SO4 and HI with a SO2 feed rate to the reactor of 50 kg/hour. Based on this value, an assumed reactor efficiency of 33%, and kinetic data from the literature, a plug flow reactor approximately 1%E2%80%9D diameter and and 12 inches long would be needed to meet the specification of the project. Because the Bunsen reaction is exothermic, heat in the amount of approximately 128,000 kJ/hr would need to be removed using a cooling jacket placed around the tubular reactor. The available literature information on Bunsen reactor design and operation, certain support equipment needed for process operation and a design that meet the specification of Matt Channon Consulting are presented.

  5. Ultra high temperature particle bed reactor design

    NASA Astrophysics Data System (ADS)

    Lazareth, Otto; Ludewig, Hans; Perkins, K.; Powell, J.

    1990-04-01

    A direct nuclear propulsion engine which could be used for a mission to Mars is designed. The main features of this reactor design are high values for I(sub sp) and very efficient cooling. This particle bed reactor consists of 37 cylindrical fuel elements embedded in a cylinder of beryllium which acts as a moderator and reflector. The fuel consists of a packed bed of spherical fissionable fuel particles. Gaseous H2 passes over the fuel bed, removes the heat, and is exhausted out of the rocket. The design was found to be neutronically critical and to have tolerable heating rates. Therefore, this particle bed reactor design is suitable as a propulsion unit for this mission.

  6. UF6 breeder reactor power plants for electric power generation

    NASA Technical Reports Server (NTRS)

    Rust, J. H.; Clement, J. D.; Hohl, F.

    1976-01-01

    The reactor concept analyzed is a U-233F6 core surrounded by a molten salt (Li(7)F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. Advantages of the gas core breeder reactor (GCBR) are as follows: (1) high efficiency; (2) simplified on-line reprocessing; (3) inherent safety considerations; (4) high breeding ratio; (5) possibility of burning all or most of the long-lived nuclear waste actinides; and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion.

  7. Design reliability assurance program for Korean next generation reactor

    SciTech Connect

    Lee, Beom-Su; Han, Jin-Kyu; Na, Jang Hwan; Yoo, Kyung Yeong

    1997-12-01

    The Korean Next Generation Reactor (KNGR) project is to develop standardized nuclear power plant design for the construction of future nuclear power plants in Korea. The main purpose of the KNGR project is to develop the advanced nuclear power plants, which enhance safety and economics significantly through the incorporation of design concepts for severe accident prevention and mitigation, supplementary passive safety concept, simplification and application of modularization and so on. For those, Probabilistic Safety Assessment (PSA) and availability study will be performed at the early stage of the design, and the Design Reliability Assurance Program (D-RAP) is applied in the development of the KNGR to ensure that the safety and availability evaluated in the PSA and availability study at the early phase of the design is maintained through the detailed design, construction, procurement and operation of the plants. This paper presents the D-RAP concept that could be applied at the stage of the basic design of the nuclear power plants, based on the models for the reference plants and/or similar plants. 4 refs., 1 fig.

  8. Westinghouse Small Modular Reactor nuclear steam supply system design

    SciTech Connect

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J.

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

  9. Liquid metal cooled nuclear reactor plant system

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  10. Clinch River Breeder Reactor Plant Project. Summary edition. 1980 technical progress report, October 1979-September 1980

    SciTech Connect

    Not Available

    1980-01-01

    This technical progress report on the CRBRP Project describes the objectives, design decisions, and major accomplishments achieved in the planning, organizing, design, and execution of the Project during the period October 1, 1979, through September 30, 1980. It is a summary of the 1980 CRBRP Technical Progress Report, which was prepared by the Advanced Reactors Division of Westinghouse Electric Corporation, the Lead Reactor Manufacturer for the Clinch River Breeder Reactor Plant Project, in fulfillment of contract requirements with the United States Department of Energy. It includes inputs from the CRBRP Architect-Engineer (Burns and Roe, Inc.), from the Constructor (Stone and Webster Engineering Corporation), and from the supporting Reactor Manufacturers (Atomics International Division of the Energy Systems Group of Rockwell International Corporation, the Advanced Reactor Systems Department of General Electric Company, and the Advanced Reactors Division of Westinghouse Electric Corporation).

  11. Membrane bio-reactor for textile wastewater treatment plant upgrading.

    PubMed

    Lubello, C; Gori, R

    2005-01-01

    Textile industries carry out several fiber treatments using variable quantities of water, from five to forty times the fiber weight, and consequently generate large volumes of wastewater to be disposed of. Membrane Bio-reactors (MBRs) combine membrane technology with biological reactors for the treatment of wastewater: micro or ultrafiltration membranes are used for solid-liquid separation replacing the secondary settling of the traditional activated sludge system. This paper deals with the possibility of realizing a new section of one existing WWTP (activated sludge + clariflocculation + ozonation) for the treatment of treating textile wastewater to be recycled, equipped with an MBR (76 l/s as design capacity) and running in parallel with the existing one. During a 4-month experimental period, a pilot-scale MBR proved to be very effective for wastewater reclamation. On average, removal efficiency of the pilot plant (93% for COD, and over 99% for total suspended solids) was higher than the WWTP ones. Color was removed as in the WWTP. Anionic surfactants removal of pilot plant was lower than that of the WWTP (90.5 and 93.2% respectively), while the BiAS removal was higher in the pilot plant (98.2 vs. 97.1). At the end cost analysis of the proposed upgrade is reported.

  12. Conceptual design of a laser-fusion power plant. Part II. Two technical options: 1. JADE reactor; 2. Heat transfer by heat pipes

    SciTech Connect

    Not Available

    1981-07-01

    A laser fusion reactor concept is described that employs liquid metal walls. The concept envisions a porous medium, called the JADE, of specific geometry lining the reactor cavity. Some advantages and disadvantages of the concept are pointed out. The possibility of using heat pipes for passive cooling in ICF reactors is discussed. Some of the problems are outlined. (MOW)

  13. Advanced burner test reactor preconceptual design report.

    SciTech Connect

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

    2008-12-16

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

  14. Assessement of Codes and Standards Applicable to a Hydrogen Production Plant Coupled to a Nuclear Reactor

    SciTech Connect

    M. J. Russell

    2006-06-01

    This is an assessment of codes and standards applicable to a hydrogen production plant to be coupled to a nuclear reactor. The result of the assessment is a list of codes and standards that are expected to be applicable to the plant during its design and construction.

  15. Preparation of plant and system design description documents

    SciTech Connect

    Not Available

    1989-01-01

    This standard prescribes the purpose, scope, organization, and content of plant design requirements (PDR) documents and system design descriptions (SDDs), to provide a unified approach to their preparation and use by a project as the principal means to establish the plant design requirements and to establish, describe, and control the individual system designs from conception and throughout the lifetime of the plant. The Electric Power Research Institute`s Advanced Light Water Reactor (LWR) Requirements Document should be considered for LWR plants.

  16. Preparation of plant and system design description documents

    SciTech Connect

    Not Available

    1989-01-01

    This standard prescribes the purpose, scope, organization, and content of plant design requirements (PDR) documents and system design descriptions (SDDs), to provide a unified approach to their preparation and use by a project as the principal means to establish the plant design requirements and to establish, describe, and control the individual system designs from conception and throughout the lifetime of the plant. The Electric Power Research Institute's Advanced Light Water Reactor (LWR) Requirements Document should be considered for LWR plants.

  17. Safety aspects of the US advanced LMR (liquid metal reactor) design

    SciTech Connect

    Pedersen, D.R.; Gyorey, G.L.; Marchaterre, J.F.; Rosen, S.; General Electric Co., San Jose, CA; Argonne National Lab., IL; USDOE Assistant Secretary for Nuclear Energy, Washington, DC )

    1989-01-01

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

  18. HYLIFE-2 inertial confinement fusion reactor design

    NASA Astrophysics Data System (ADS)

    Moir, Ralph W.

    1990-10-01

    The HYLIFE-II inertial fusion power plant design study uses a liquid fall, in the form of jets to protect the first structural wall from neutron damage, x rays, and blast to provide a 30-y lifetime. HYLIFE-I used liquid lithium. HYLIFE-II avoids the fire hazard of lithium by using a molten salt composed of fluorine, lithium, and beryllium (Li2BeF4) called Flibe. Access for heavy-ion beams is provided. Calculations for assumed heavy-ion beam performance show a nominal gain of 70 at 5 MJ producing 350 MJ, about 5.2 times less yield than the 1.8 GJ from a driver energy of 4.5 MJ with gain of 400 for HYLIFE-I. The nominal 1 GWe of power can be maintained by increasing the repetition rate by a factor of about 5.2, from 1.5 to 8 Hz. A higher repetition rate requires faster re-establishment of the jets after a shot, which can be accomplished in part by decreasing the jet fall height and increasing the jet flow velocity. Multiple chambers may be required. In addition, although not considered for HYLIFE-I, there is undoubtedly liquid splash that must be forcibly cleared because gravity is too slow, especially at high repetition rates. Splash removal can be accomplished by either pulsed or oscillating jet flows. The cost of electricity is estimated to be 0.09 $/kW h in constant 1988 dollars, about twice that of future coal and light water reactor nuclear power. The driver beam cost is about one-half the total cost.

  19. Advanced Neutron Sources: Plant Design Requirements

    SciTech Connect

    Not Available

    1990-07-01

    The Advanced Neutron Source (ANS) is a new, world class facility for research using hot, thermal, cold, and ultra-cold neutrons. At the heart of the facility is a 350-MW{sub th}, heavy water cooled and moderated reactor. The reactor is housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides fans out into a large guide hall, housing about 30 neutron research stations. Office, laboratory, and shop facilities are included to provide a complete users facility. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory at the end of the decade. This Plant Design Requirements document defines the plant-level requirements for the design, construction, and operation of the ANS. This document also defines and provides input to the individual System Design Description (SDD) documents. Together, this Plant Design Requirements document and the set of SDD documents will define and control the baseline configuration of the ANS.

  20. United States Department of Energy`s reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    SciTech Connect

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage.

  1. Work Breakdown Structure and Plant/Equipment Designation System Numbering Scheme for the High Temperature Gas- Cooled Reactor (HTGR) Component Test Capability (CTC)

    SciTech Connect

    Jeffrey D Bryan

    2009-09-01

    This white paper investigates the potential integration of the CTC work breakdown structure numbering scheme with a plant/equipment numbering system (PNS), or alternatively referred to in industry as a reference designation system (RDS). Ideally, the goal of such integration would be a single, common referencing system for the life cycle of the CTC that supports all the various processes (e.g., information, execution, and control) that necessitate plant and equipment numbers be assigned. This white paper focuses on discovering the full scope of Idaho National Laboratory (INL) processes to which this goal might be applied as well as the factors likely to affect decisions about implementation. Later, a procedure for assigning these numbers will be developed using this white paper as a starting point and that reflects the resolved scope and outcome of associated decisions.

  2. 78 FR 17945 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-03-25

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant Designs; Notice of Meeting The ACRS Subcommittee on Future Plant Designs will hold a meeting on April 9... based licensing framework for the Next Generation Nuclear Plant (NGNP). The Subcommittee will...

  3. 77 FR 74698 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-17

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant Designs; Notice of Meeting The ACRS Subcommittee on Future Plant Designs will hold a meeting on January 17... Generation Nuclear Plant (NGNP) fuel and source term research and development of risk-informed...

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

    SciTech Connect

    Not Available

    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.

  5. New Generation Nuclear Plant (NGNP) Project, Preliminary Point Design

    SciTech Connect

    F. H. Southworth; P. E. MacDonald; A. M. Baxter; P. D. Bayless; J. M. Bolin; H. D. Gougar; R. L. Moore; A. M. Ougouag; M. B. Richards; R. L. Sant; J. W. Sterbentz; W. K. Terry

    2004-03-01

    This paper provides a preliminary assessment of two possible versions of the Next Generation Nuclear Plant (NGNP), a prismatic fuel type helium gas-cooled reactor and a pebblebed fuel helium gas reactor. Both designs will meet the three basic requirements that have been set for the NGNP: a coolant outlet temperature of 1000 C, passive safety, and a total power output consistent with that expected for commercial high-temperature gas-cooled reactors.

  6. 75 FR 58448 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee On Future Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-24

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee On Future Plant Designs The ACRS Subcommittee on Future Plant Designs will hold a meeting on October 21, 2010, at 11545... Subcommittee will review current Design Acceptance Criteria associated with Digital Instrumentation and...

  7. Neutron transport analysis for nuclear reactor design

    DOEpatents

    Vujic, J.L.

    1993-11-30

    Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values. 28 figures.

  8. Neutron transport analysis for nuclear reactor design

    DOEpatents

    Vujic, Jasmina L.

    1993-01-01

    Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values.

  9. Designing a SCADA system simulator for fast breeder reactor

    NASA Astrophysics Data System (ADS)

    Nugraha, E.; Abdullah, A. G.; Hakim, D. L.

    2016-04-01

    SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

  10. High Efficiency Nuclear Power Plants using Liquid Fluoride Thorium Reactor Technology

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITS of 950 K and 1200 K are presented. Power plant performance data were obtained for TITS ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo -generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  11. High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  12. Design and performance of subgrade biogeochemical reactors.

    PubMed

    Gamlin, Jeff; Downey, Doug; Shearer, Brad; Favara, Paul

    2017-02-18

    Subgrade biogeochemical reactors (SBGRs), also commonly referred to as in situ bioreactors, are a unique technology for treatment of contaminant source areas and groundwater plume hot spots. SBGRs have most commonly been configured for enhanced reductive dechlorination (ERD) applications for chlorinated solvent treatment. However, they have also been designed for other contaminant classes using alternative treatment media. The SBGR technology typically consists of removal of contaminated soil via excavation or large-diameter augers, and backfill of the soil void with gravel and treatment amendments tailored to the target contaminant(s). In most cases SBGRs include installation of infiltration piping and a low-flow pumping system (typically solar-powered) to recirculate contaminated groundwater through the SBGR for treatment. SBGRs have been constructed in multiple configurations, including designs capable of meeting limited access restrictions at heavily industrialized sites, and at sites with restrictions on surface disturbance due to sensitive species or habitat issues. Typical performance results for ERD applications include 85 to 90 percent total molar reduction of chlorinated volatile organic compounds (CVOCs) near the SBGR and rapid clean-up of adjacent dissolved contaminant source areas. Based on a review of the literature and CH2M's field-scale results from over a dozen SBGRs with a least one year of performance data, important site-specific design considerations include: 1) hydraulic residence time should be long enough for sufficient treatment but not too long to create depressed pH and stagnant conditions (e.g., typically between 10 and 60 days), 2) reactor material should balance appropriate organic mulch as optimal bacterial growth media along with other organic additives that provide bioavailable organic carbon, 3) a variety of native bacteria are important to the treatment process, and 4) biologically mediated generation of iron sulfides along with

  13. Applications of high-strength concrete to the development of the prestressed concrete reactor vessel (PCRV) design for an HTGR-SC/C plant

    SciTech Connect

    Naus, D.J.

    1984-01-01

    The PCRV research and development program at ORNL consists of generic studies to provide technical support for ongoing PCRV-related studies, to contribute to the technological data base, and to provide independent review and evaluation of the relevant technology. Recent activities under this program have concentrated on the development of high-strength concrete mix designs for the PCRV of a 2240 MW(t) HTGR-SC/C plant, and the testing of models to both evaluate the behavior of high-strength concretes (plain and fibrous) and to develop model testing techniques. A test program to develop and evaluate high-strength (greater than or equal to 63.4 MPa) concretes utilizing materials from four sources which are in close proximity to potential sites for an HTGR plant is currently under way. The program consists of three phases. Phase I involves an evaluation of the cement, fly ash, admixtures and aggregate materials relative to their capability to produce concretes having the desired strength properties. Phase II is concerned with the evaluation of the effects of elevated temperatures (less than or equal to 316/sup 0/C) on the strength properties of mixes selected for detailed evaluation. Phase III involves a determination of the creep characteristics and thermal properties of the selected mixes. An overview of each of these phases is presented as well as results obtained to date under Phase I which is approximately 75% completed.

  14. A Methodology for the Neutronics Design of Space Nuclear Reactors

    SciTech Connect

    King, Jeffrey C.; El-Genk, Mohamed S.

    2004-02-04

    A methodology for the neutronics design of space power reactors is presented. This methodology involves balancing the competing requirements of having sufficient excess reactivity for the desired lifetime, keeping the reactor subcritical at launch and during submersion accidents, and providing sufficient control over the lifetime of the reactor. These requirements are addressed by three reactivity values for a given reactor design: the excess reactivity at beginning of mission, the negative reactivity at shutdown, and the negative reactivity margin in submersion accidents. These reactivity values define the control worth and the safety worth in submersion accidents, used for evaluating the merit of a proposed reactor type and design. The Heat Pipe-Segmented Thermoelectric Module Converters space reactor core design is evaluated and modified based on the proposed methodology. The final reactor core design has sufficient excess reactivity for 10 years of nominal operation at 1.82 MW of fission power and is subcritical at launch and in all water submersion accidents.

  15. An Overview of Reactor Concepts, a Survey of Reactor Designs.

    DTIC Science & Technology

    1985-02-01

    reactor core at the top and discharged at the Dotton while the reactor is in operation. The discharged fuel can then b inspected to see if it can De used...Public Affairs Office and is releasaole to the National Technical Information Services (NTIS). At NTIS, it will be available to the general public...addressee, please notify AFWL/NTYN, Kirtland AFB, NN 87117 to help us maintain a current mailing list. This report has been reviewed and is approved

  16. Conceptual design study of JSFR reactor building

    SciTech Connect

    Yamamoto, T.; Katoh, A.; Chikazawa, Y.; Ohya, T.; Iwasaki, M.; Hara, H.; Akiyama, Y.

    2012-07-01

    Japan Sodium-cooled Fast Reactor (JSFR) is planning to adopt the new concepts of reactor building. One is that the steel plate reinforced concrete is adopted for containment vessel and reactor building. The other is the advanced seismic isolation system. This paper describes the detail of new concepts for JSFR reactor building and engineering evaluation of the new concepts. (authors)

  17. Study of hydrogen generation plant coupled to high temperature gas cooled reactor

    NASA Astrophysics Data System (ADS)

    Brown, Nicholas Robert

    Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Several thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. Eight unique case studies are performed based on a thorough literature review of possible events. The case studies are: (1) feed flow failure from one section of the chemical plant to another, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without SCRAM, (6) total failure of the chemical plant, (7) parametric study of the temperature in an individual reaction chamber, and (8) control rod insertion in the nuclear reactor. Various parametric

  18. Analysis of reactor trips originating in balance of plant systems

    SciTech Connect

    Stetson, F.T.; Gallagher, D.W.; Le, P.T.; Ebert, M.W. )

    1990-09-01

    This report documents the results of an analysis of balance-of-plant (BOP) related reactor trips at commercial US nuclear power plants of a 5-year period, from January 1, 1984, through December 31, 1988. The study was performed for the Plant Systems Branch, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission. The objectives of the study were: to improve the level of understanding of BOP-related challenges to safety systems by identifying and categorizing such events; to prepare a computerized data base of BOP-related reactor trip events and use the data base to identify trends and patterns in the population of these events; to investigate the risk implications of BOP events that challenge safety systems; and to provide recommendations on how to address BOP-related concerns in regulatory context. 18 refs., 2 figs., 27 tabs.

  19. Inertial Fusion Energy Reactor Design Studies: Prometheus-L, Prometheus-H. Volume 3, Final report

    SciTech Connect

    Waganer, L.M.; Driemeyer, D.E.; Lee, V.D.

    1992-03-01

    This report contains a review of design studies for inertial confinement reactor. This third of three three volumes discusses the following topics: Driver system definition; vacuum system; fuel processing systems (FPS); cavity design and analysis; heat transport and thermal energy conversion; balance of plant systems; remote maintenance systems; safety and environment; economics; and comparison of IFE designs.

  20. Nuclear Technology Series. Nuclear Reactor (Plant) Operator Trainee. A Suggested Program Planning Guide. Revised June 80.

    ERIC Educational Resources Information Center

    Center for Occupational Research and Development, Inc., Waco, TX.

    This program planning guide for a two-year postsecondary nuclear reactor (plant) operator trainee program is designed for use with courses 1-16 of thirty-five in the Nuclear Technology Series. The purpose of the guide is to describe the nuclear power field and its job categories for specialists, technicians and operators; and to assist planners,…

  1. Design and evaluation of experimental ceramic automobile thermal reactors

    NASA Technical Reports Server (NTRS)

    Stone, P. L.; Blankenship, C. P.

    1974-01-01

    The paper summarizes the results obtained in an exploratory evaluation of ceramics for automobile thermal reactors. Candidate ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance, lasting 1100 hours in engine dynamometer tests and for more than 38,600 kilimeters (24,000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

  2. Design and evaluation of experimental ceramic automobile thermal reactors

    NASA Technical Reports Server (NTRS)

    Stone, P. L.; Blankenship, C. P.

    1974-01-01

    The results obtained in an exploratory evaluation of ceramics for automobile thermal reactors are summarized. Candidate ceramic materials were evaluated in several reactor designs by using both engine-dynamometer and vehicle road tests. Silicon carbide contained in a corrugated-metal support structure exhibited the best performance, lasting 1100 hr in engine-dynamometer tests and more than 38,600 km (24000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as those containing silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

  3. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    SciTech Connect

    Teofilo, V.L.; Leonard, B.R. Jr.; Aase, D.T.

    1980-09-01

    The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant has been performed. The tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb/sub 3/Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs have been made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis has been made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered.

  4. 76 FR 5220 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-28

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant Designs The ACRS Subcommittee on Future Plant Designs will hold a meeting on February 9, 2011, at...

  5. 76 FR 64123 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-10-17

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant Designs; Notice of Meeting The ACRS Subcommittee on Future Plant Designs will hold a meeting on November...

  6. 76 FR 16016 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-22

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant Designs The ACRS Subcommittee on Future Plant Designs will hold a meeting on April 5, 2011, at...

  7. HVAC [Heating, Ventilation and Air Conditioning] subsystem design description: 4 x 350 MW(t) Modular HTGR [High-Temperature Gas-Cooled Reactor] Plant

    SciTech Connect

    1986-06-01

    The HVAC system is a subsystem within the Mechanical Services Group (MSG). The HVAC system for the 4 x 350 MW(t) Modular HTGR Plant presently consists of ten, nonsafety-related subsystems located in the Nuclear Island (NI) and Energy Conversion Area (ECA) of the plant.

  8. Conceptual design of the advanced marine reactor MRX

    NASA Astrophysics Data System (ADS)

    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.

  9. Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

    NASA Astrophysics Data System (ADS)

    Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

    2006-01-01

    A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

  10. Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

    SciTech Connect

    Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

    2006-01-20

    A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

  11. GAS-PASS/H : a simulation code for gas reactor plant systems.

    SciTech Connect

    Vilim, R. B.; Mertyurek, U.; Cahalan, J. E.; Nuclear Engineering Division; Texas A&M Univ.

    2004-01-01

    A simulation code for gas reactor plant systems has been developed. The code is intended for use in safety analysis and control studies for Generation-IV reactor concepts. We developed it anticipating an immediate application to direct cycle gas reactors. By programming in flexibility as to how components can be configured, we believe the code can be adapted for the indirect-cycle gas reactors relatively easy. The use of modular components and a general purpose equation solver allows for this. There are several capabilities that are included for investigating issues associated with direct cycle gas reactors. Issues include the safety characteristics of single shaft plants during coastdown and transition to shutdown heat removal following unprotected accidents, including depressurization, and the need for safety grade control systems. Basic components provided include turbine, compressor, recuperator, cooler, bypass valve, leak, accumulator, containment, and flow junction. The code permits a more rapid assessment of design concepts than is achievable using RELAP. RELAP requires detail beyond what is necessary at the design scoping stage. This increases the time to assemble an input deck and tends to make the code slower running. The core neutronics and decay heat models of GAS-PASS/H are taken from the liquid-metal version of MINISAS. The ex-reactor component models were developed from first principles. The network-based method for assembling component models into a system uses a general nonlinear solver to find the solution to the steady-state equations. The transient time-differenced equations are solved implicitly using the same solver. A direct cycle gas reactor is modeled and a loss of generator load transient is simulated for this reactor. While normally the reactor is scrammed, the plant safety case will require analysis of this event with failure of various safety systems. Therefore, we simulated the loss of load transient with a combined failure of the

  12. Advanced Neutron Source: Plant Design Requirements

    SciTech Connect

    Not Available

    1990-07-01

    The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS.

  13. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    SciTech Connect

    Pope, R B; Diggs, J M

    1982-04-01

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

  14. Transpiring wall supercritical water oxidation test reactor design report

    SciTech Connect

    Haroldsen, B.L.; Ariizumi, D.Y.; Mills, B.E.; Brown, B.G.; Rousar, D.C.

    1996-02-01

    Sandia National Laboratories is working with GenCorp, Aerojet and Foster Wheeler Development Corporation to develop a transpiring wall supercritical water oxidation reactor. The transpiring wall reactor promises to mitigate problems of salt deposition and corrosion by forming a protective boundary layer of pure supercritical water. A laboratory scale test reactor has been assembled to demonstrate the concept. A 1/4 scale transpiring wall reactor was designed and fabricated by Aerojet using their platelet technology. Sandia`s Engineering Evaluation Reactor serves as a test bed to supply, pressurize and heat the waste; collect, measure and analyze the effluent; and control operation of the system. This report describes the design, test capabilities, and operation of this versatile and unique test system with the transpiring wall reactor.

  15. Reactor Design from a Stability Viewpoint.

    ERIC Educational Resources Information Center

    Perlmutter, D. D.

    1978-01-01

    This course uses stability as a central theme around which to organize a wide range of reactor concerns. This approach brings together the subject matter of catalyst particles with that of well-stirred vessels and tubular reactor geometry. (Author/BB)

  16. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    SciTech Connect

    Hongbin Zhang; Haihua Zhao

    2009-05-01

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

  17. Makeup water treatment and auxiliary boiler building structural design description: 4 x 350 MW(t) Modular HTGR [High-Temperature Gas-Cooled Reactor] Plant

    SciTech Connect

    1986-06-01

    The Makeup Water Treatment and Auxiliary Boiler Building (MWABB) is a grade-founded, single-story, steel-framed structure with insulated sheet metal exterior walls and roof decking. It houses the electrically-heated auxiliary boiler and related equipment, and the Raw Water Treatment System. The Makeup Water Treatment and Auxiliary Boiler building is located adjacent to the Maintenance Building in the Energy Conversion Area of the plant.

  18. Space station prototype Sabatier reactor design verification testing

    NASA Technical Reports Server (NTRS)

    Cusick, R. J.

    1974-01-01

    A six-man, flight prototype carbon dioxide reduction subsystem for the SSP ETC/LSS (Space Station Prototype Environmental/Thermal Control and Life Support System) was developed and fabricated for the NASA-Johnson Space Center between February 1971 and October 1973. Component design verification testing was conducted on the Sabatier reactor covering design and off-design conditions as part of this development program. The reactor was designed to convert a minimum of 98 per cent hydrogen to water and methane for both six-man and two-man reactant flow conditions. Important design features of the reactor and test conditions are described. Reactor test results are presented that show design goals were achieved and off-design performance was stable.

  19. Plant design: Integrating Plant and Equipment Models

    SciTech Connect

    Sloan, David; Fiveland, Woody; Zitney, S.E.; Osawe, Maxwell

    2007-08-01

    Like power plant engineers, process plant engineers must design generating units to operate efficiently, cleanly, and profitably despite fluctuating costs for raw materials and fuels. To do so, they increasingly create virtual plants to enable evaluation of design concepts without the expense of building pilot-scale or demonstration facilities. Existing computational models describe an entire plant either as a network of simplified equipment models or as a single, very detailed equipment model. The Advanced Process Engineering Co-Simulator (APECS) project (Figure 5) sponsored by the U.S. Department of Energy's National Energy Technology Laboratory (NETL) seeks to bridge the gap between models by integrating plant modeling and equipment modeling software. The goal of the effort is to provide greater insight into the performance of proposed plant designs. The software integration was done using the process-industry standard CAPE-OPEN (Computer Aided Process Engineering–Open), or CO interface. Several demonstration cases based on operating power plants confirm the viability of this co-simulation approach.

  20. Conceptual design of module fast reactor of ultimate safety cooled by lead-bismuth alloy

    SciTech Connect

    Myasnikov, V.O.; Stekolnikov, V.V.; Stepanov, V.S.; Gorshkov, V.T.; Kulikov, M.L.; Shulyndin, V.A.; Gromov, B.F.; Kalashnikov, A.G.; Pashkin, Yu.G.

    1993-12-31

    During past time all basic problems arisen during working-out of NPP with lead-bismuth coolant were solved: physics and thermal physics of the cores, heat transfer and hydrodynamics, corrosion resistance of the structural materials and coolant technology, radiation and nuclear safety, investigations of emergency situations, development of fuel elements and absorbing elements of the reactor, equipment of the primary circuit and other circuits. A powerful experimental base equpped by unique rigs is made. A series of ship and test NPP has been constructed whereat repair of the plants and reactor refuelling are developed. Highly-skilled groups of investigators, designers and operation personnel capable of performing the development of the reactor plant with MFR within short terms have been formed. In this case MFR with lead-bismuth coolant may become the initial step in development of large-scale nuclear power engineering with fast reactors cooled by liquid lead.

  1. ANALYSIS OF A HIGH TEMPERATURE GAS-COOLED REACTOR POWERED HIGH TEMPERATURE ELECTROLYSIS HYDROGEN PLANT

    SciTech Connect

    M. G. McKellar; E. A. Harvego; A. M. Gandrik

    2010-11-01

    An updated reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322°C and 750°C, respectively. The reactor heat is used to produce heat and electric power to the HTE plant. A Rankine steam cycle with a power conversion efficiency of 44.4% was used to provide the electric power. The electrolysis unit used to produce hydrogen includes 1.1 million cells with a per-cell active area of 225 cm2. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 42.8% at a hydrogen production rate of 1.85 kg/s (66 million SCFD) and an oxygen production rate of 14.6 kg/s (33 million SCFD). An economic analysis of this plant was performed with realistic financial and cost estimating The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.03/kg of hydrogen was calculated assuming an internal rate of return of 10% and a debt to equity ratio of 80%/20% for a reactor cost of $2000/kWt and $2.41/kg of hydrogen for a reactor cost of $1400/kWt.

  2. Neutronic Reactor Design to Reduce Neutron Loss

    DOEpatents

    Miles, F. T.

    1961-05-01

    A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall. The wall is surrounded by successive layers of pure fertile material and moderator containing fertile material. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. Since the steel has a smaller capture cross section for the fast neutrons, greater nunnbers of neutrons will pass into the blanket, thereby increasing the over-all efficiency of the reactor. (AEC)

  3. NEUTRONIC REACTOR DESIGN TO REDUCE NEUTRON LOSS

    DOEpatents

    Mills, F.T.

    1961-05-01

    A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall which is surrounded by successive layers of pure fertile material and fertile material having moderator. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. As the steel has a smaller capture cross-section for the fast neutrons, then greater numbers of the neutrons will pass into the blanket thereby increasing the over-all efficiency of the reactor.

  4. Reactor design and integration into a nuclear electric spacecraft

    NASA Technical Reports Server (NTRS)

    Phillips, W. M.; Koenig, D. R.

    1978-01-01

    One of the well-defined applications for nuclear power in space is nuclear electric propulsion (NEP). Mission studies have identified the optimum power level (400 kWe). A single Shuttle launch requirement and science-package integration have added additional constraints to the design. A reactor design which will meet these constraints has been studied. The reactor employs 90 fuel elements, each heat pipe cooled. Reactor control is obtained with BeO/B4C drums in a BeO reflector. The balance of the spacecraft is shielded from the reactor with LiH. Power conditioning and reactor control drum drives are located behind the LiH with the power conditioning. Launch safety, mechanical design and integration with the power conversion subsystem are discussed.

  5. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    SciTech Connect

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  6. High-Temperature Gas-Cooled Test Reactor Point Design

    SciTech Connect

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

    2016-04-01

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

  7. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have

  8. Animal Guts as Ideal Reactors: An Open-Ended Project for a Course in Kinetics and Reactor Design.

    ERIC Educational Resources Information Center

    Carlson, Eric D.; Gast, Alice P.

    1998-01-01

    Presents an open-ended project tailored for a senior kinetics and reactor design course in which basic reactor design equations are used to model the digestive systems of several animals. Describes the assignment as well as the results. (DDR)

  9. Fast Reactor Alternative Studies: Effects of Transuranic Groupings on Metal and Oxide Sodium Fast Reactor Designs

    SciTech Connect

    R. Ferrer; M. Asgari; S. Bays; B. Forget

    2007-09-01

    A 1000 MWth commercial-scale Sodium Fast Reactor (SFR) design with a conversion ratio (CR) of 0.50 was selected in this study to perform perturbations on the external feed coming from Light Water Reactor Spent Nuclear Fuel (LWR SNF) and separation groupings in the reprocessing scheme. A secondary SFR design with a higher conversion ratio (CR=0.75) was also analyzed as a possible alternative, although no perturbations were applied to this model.

  10. Light water reactor program

    SciTech Connect

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  11. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    SciTech Connect

    Woo, H.H.; Lu, S.C.

    1981-09-15

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design.

  12. Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen Production Plant

    SciTech Connect

    E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

    2008-08-01

    A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540°C and 900°C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating-current, AC, to direct-current, DC, conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

  13. VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

    SciTech Connect

    Doshi, P.K.; DeMario, E.E.; Knott, R.P.

    1993-12-31

    Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR`s) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design.

  14. DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

    SciTech Connect

    Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

    2003-11-12

    The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

  15. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    SciTech Connect

    Professor Neill Todreas

    2001-10-01

    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team indicates

  16. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  17. Annealing the reactor vessel at the Palisades Plant

    SciTech Connect

    Fenech, R.A.

    1996-03-01

    In the way of background, Palisades was licensed in 1967 and went commercial in 1971. Jumping to two years ago, we faced at that time three issues that challenged our ability to operate to end-of-license, which would be 2007 without any extensions. The three items were regulatory performance, economic performance, and reactor vessel embrittlement. We had not been operating the plant with the kind of conservative decisions and with the kind of safety margins that one is expected to operate a plant in the United States at this time. Our economic performance was not satisfactory in that our capacity factor was low and our costs high. In the area of reactor vessel embrittlement, our analysis showed that we would reach the NRC screening criteria for embrittlement in the year 2004. Over the last two years, we have made significant improvements in the first two areas. Our decision-making has changed. Our performance, especially over the last year and a half, has been excellent. In addition, we have gotten our capacity factors up and our costs under control. Clearly, sustained performance is what is going to carry the day but from what we can see and from where we are, we are in more of a maintenance-of-performance than in a turn-around situation. On the other hand, in the area of reactor vessel embrittlement, about a year and a half ago we had a bit of a setback. We had taken material from retired steam generators that had welds identical to the welds in our reactor vessel. When we analyzed the welds from our steam generators, we were given some surprises about the chemistry makeup. When we applied the new information to our analysis, we changed the date on which we would reach our screening criteria from 2004 to late 1999.

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

    SciTech Connect

    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

  19. Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors

    SciTech Connect

    Holbrook, Mark; Kinsey, Jim

    2015-03-01

    In July 2013, the US Department of Energy (DOE) and US Nuclear Regulatory Commission (NRC) established a joint initiative to address a key portion of the licensing framework essential to advanced (non-light water) reactor technologies. The initiative addressed the “General Design Criteria for Nuclear Power Plants,” Appendix A to10 Code of Federal Regulations (CFR) 50, which were developed primarily for light water reactors (LWRs), specific to the needs of advanced reactor design and licensing. The need for General Design Criteria (GDC) clarifications in non-LWR applications has been consistently identified as a concern by the industry and varied stakeholders and was acknowledged by the NRC staff in their 2012 Report to Congress1 as an area for enhancement. The initiative to adapt GDC requirements for non-light water advanced reactor applications is being accomplished in two phases. Phase 1, managed by DOE, consisted of reviews, analyses and evaluations resulting in recommendations and deliverables to NRC as input for NRC staff development of regulatory guidance. Idaho National Laboratory (INL) developed this technical report using technical and reactor technology stakeholder inputs coupled with analysis and evaluations provided by a team of knowledgeable DOE national laboratory personnel with input from individual industry licensing consultants. The DOE national laboratory team reviewed six different classes of emerging commercial reactor technologies against 10 CFR 50 Appendix A GDC requirements and proposed guidance for their adapted use in non-LWR applications. The results of the Phase 1 analysis are contained in this report. A set of draft Advanced Reactor Design Criteria (ARDC) has been proposed for consideration by the NRC in the establishment of guidance for use by non-LWR designers and NRC staff. The proposed criteria were developed to preserve the underlying safety bases expressed by the original GDC, and recognizing that advanced reactors may take

  20. Regulatory Review of the Digital Plant Protection System for Advanced Power Reactor 1400

    SciTech Connect

    Kim, DAI. I.; Ji, S.H.; Park, H.S.; Kim, B.R.; Kang, Y.D.; Oh, S.H.

    2002-07-01

    This paper presents the evaluation result and the regulatory approach of digital plant protection system (DPPS) for Advanced Power Reactor (APR-1400). Firstly, we discuss the issue associated with the integration of bistable processor (BP) and local coincidence logic processor (LCLP) as one of design changes over digital plant protection system. Secondly, regulatory approach is presented on the safety classification and the independence of the soft controller to be installed in digital engineered safety features actuation system (DESFAS). Finally, hardwired back up systems against common mode failure of a digital system and the safety classification of Remote Shutdown Panel (RSP) are described. (authors)

  1. The Next Generation Nuclear Plant Graphite Creep Experiment Irradiation in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep 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 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 graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of

  2. Design criteria for prestressed concrete reactor vessels for high-temperature reactors

    SciTech Connect

    Elter, C.; Becker, G.

    1982-11-01

    For the design and construction of prestressed concrete reactor vessels, data on loading, construction materials, and safety factors are required. A description is given of the design conditions according to the current state of technology in the Federal Republic of Germany. Special consideration is given to the allowable stresses and an appropriate proposal for such stresses is suggested.

  3. Introduction to Chemical Engineering Reactor Analysis: A Web-Based Reactor Design Game

    ERIC Educational Resources Information Center

    Orbey, Nese; Clay, Molly; Russell, T.W. Fraser

    2014-01-01

    An approach to explain chemical engineering through a Web-based interactive game design was developed and used with college freshman and junior/senior high school students. The goal of this approach was to demonstrate how to model a lab-scale experiment, and use the results to design and operate a chemical reactor. The game incorporates both…

  4. Shield Design for a Space Based Vapor Core Reactor

    SciTech Connect

    Knight, Travis; Anghaie, Samim

    2002-07-01

    Innovative shielding strategies were sought to reduce the mass of the required shielding for a space based vapor core reactor system with magnetohydrodynamic energy conversion. Gamma-rays directly resultant from fission were found to play no role in the dose rate, while secondary gamma-rays from fission neutron interactions were the dominant contributor to the dose rate. Hydrogen containing materials such as polyethylene were utilized to provide shielding of both radiation from the reactor complex and also solar and galactic cosmic radiation. This shield design was found to contribute 0.125 kg/kWe to the baseline vapor core reactor system specific mass. (authors)

  5. SABR fusion-fission hybrid transmutation reactor design concept

    NASA Astrophysics Data System (ADS)

    Stacey, Weston

    2009-11-01

    A conceptual design has been developed for a sub-critical advanced burner reactor (SABR) consisting of i) a sodium cooled fast reactor fueled with the transuranics (TRU) from spent nuclear fuel, and ii) a D-T tokamak fusion neutron source based on ITER physics and technology. Subcritical operation enables more efficient transmutation fuel cycles in TRU fueled reactors (without compromising safety), which may be essential for significant reduction in high-level waste repository requirements. ITER will serve as the prototype for the fusion neutron source, which means SABRs could be implemented to help close the nuclear fuel cycle during the 2^nd quarter of the century.

  6. Design and testing of integrated circuits for reactor protection channels

    SciTech Connect

    Battle, R.E.; Vandermolen, R.I.; Jagadish, U.; Swail, B.K.; Naser, J.

    1995-06-01

    Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. The purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing. A demonstration model for protection system of PWR reactor has been designed and built.

  7. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  8. Rocklick Plant designed for flexibility

    SciTech Connect

    Mason, R.H.

    1987-10-01

    Eastern Associated Coal Corp. constructed the 1,200-tph Rocklick Preparation Plant to complement its new system of production planning. Rocklick is designed to clean coal from all of Eastern Associated's mines and contractors working in the Pond Fork Valley of southwestern West Virginia. The Rocklick Plant can process and blend about three million tpy to meet customer specifications. The plant can ship coal by barge or rail to any market. Thus the plant is a key link in marketing Eastern's low-sulfur steam coal. To fulfill a wide range of customer specifications the Rocklick Plant is equipped to provide fully washed, high-volatile steam coal; partially washed, high-volatile steam coal; and blends of fully washed and raw coal. This article explains how the plant operates.

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

    SciTech Connect

    Trond Bjornard; John Hockert

    2011-08-01

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

  10. Design of slurry reactor for indirect liquefaction applications

    SciTech Connect

    Prakash, A.; Bendale, P.G.

    1991-01-01

    The objective of this project is to design and model a conceptual slurry reactor for two indirect liquefaction applications; (1) production of methanol and (2) production of hydrocarbon fuels via Fischer-Tropsch route. A slurry reactor is defined here as a three-phase bubble column reactor using a fine catalyst particle suspension in a high molecular weight liquid. The feed gas is introduced through spargers. It then bubbles through the column providing the agitation necessary for catalyst suspension and mass transfer. The reactor models for the two processes have been formulated using computer simulation. Process data, kinetic and thermodynamic data, heat and mass transfer data and hydrodynamic data have been used in the mathematical models to describe the slurry reactor for each of the two processes. Available data from process development units and demonstration units were used to test and validate the models. Commercial size slurry reactors for methanol and Fischer-Tropsch synthesis were sized using reactor models developed in this report.

  11. Design of slurry reactor for indirect liquefaction applications. Final report

    SciTech Connect

    Prakash, A.; Bendale, P.G.

    1991-12-31

    The objective of this project is to design and model a conceptual slurry reactor for two indirect liquefaction applications; (1) production of methanol and (2) production of hydrocarbon fuels via Fischer-Tropsch route. A slurry reactor is defined here as a three-phase bubble column reactor using a fine catalyst particle suspension in a high molecular weight liquid. The feed gas is introduced through spargers. It then bubbles through the column providing the agitation necessary for catalyst suspension and mass transfer. The reactor models for the two processes have been formulated using computer simulation. Process data, kinetic and thermodynamic data, heat and mass transfer data and hydrodynamic data have been used in the mathematical models to describe the slurry reactor for each of the two processes. Available data from process development units and demonstration units were used to test and validate the models. Commercial size slurry reactors for methanol and Fischer-Tropsch synthesis were sized using reactor models developed in this report.

  12. OSIRIS and SOMBRERO Inertial Fusion Power Plant Designs, Volume 2: Designs, Assessments, and Comparisons

    SciTech Connect

    Meier, W. R.; Bieri, R. L.; Monsler, M. J.; Hendricks, C. D.; Laybourne, P.; Shillito, K. R.

    1992-03-01

    This is a comprehensive design study of two Inertial Fusion Energy (IFE) electric power plants. Conceptual designs are presented for a fusion reactor (called Osiris) using an induction-linac heavy-ion beam driver, and another (called SOMBRERO) using a KrF laser driver. The designs covered all aspects of IFE power plants, including the chambers, heat transport and power conversion systems, balance-of-plant facilities, target fabrication, target injection and tracking, as well as the heavy-ion and KrF drivers. The point designs were assessed and compared in terms of their environmental & safety aspects, reliability and availability, economics, and technology development needs.

  13. 78 FR 28896 - Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-16

    ... COMMISSION Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components... Combinations for Metal Primary Reactor Containment System Components,'' in which there are no substantive... loading combinations for metal primary reactor containment system components. ADDRESSES: Please refer...

  14. Design of megawatt power level heat pipe reactors

    SciTech Connect

    Mcclure, Patrick Ray; Poston, David Irvin; Dasari, Venkateswara Rao; Reid, Robert Stowers

    2015-11-12

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

  15. Designed porosity materials in nuclear reactor components

    DOEpatents

    Yacout, A. M.; Pellin, Michael J.; Stan, Marius

    2016-09-06

    A nuclear fuel pellet with a porous substrate, such as a carbon or tungsten aerogel, on which at least one layer of a fuel containing material is deposited via atomic layer deposition, and wherein the layer deposition is controlled to prevent agglomeration of defects. Further, a method of fabricating a nuclear fuel pellet, wherein the method features the steps of selecting a porous substrate, depositing at least one layer of a fuel containing material, and terminating the deposition when the desired porosity is achieved. Also provided is a nuclear reactor fuel cladding made of a porous substrate, such as silicon carbide aerogel or silicon carbide cloth, upon which layers of silicon carbide are deposited.

  16. Design of Aerosol Coating Reactors: Precursor Injection

    PubMed Central

    Buesser, Beat; Pratsinis, Sotiris E.

    2013-01-01

    Particles are coated with thin shells to facilitate their processing and incorporation into liquid or solid matrixes without altering core particle properties (coloristic, magnetic, etc.). Here, computational fluid and particle dynamics are combined to investigate the geometry of an aerosol reactor for continuous coating of freshly-made titanium dioxide core nanoparticles with nanothin silica shells by injection of hexamethyldisiloxane (HMDSO) vapor downstream of TiO2 particle formation. The focus is on the influence of HMDSO vapor jet number and direction in terms of azimuth and inclination jet angles on process temperature and coated particle characteristics (shell thickness and fraction of uncoated particles). Rapid and homogeneous mixing of core particle aerosol and coating precursor vapor facilitates synthesis of core-shell nanoparticles with uniform shell thickness and high coating efficiency (minimal uncoated core and free coating particles). PMID:23658471

  17. Design-Only Conceptual Design Report: Plutonium Immobilization Plant

    SciTech Connect

    DiSabatino, A.; Loftus, D.

    1999-01-01

    This design-only conceptual design report was prepared to support a funding request by the Department of Energy Office of Fissile Materials Disposition for engineering and design of the Plutonium Immobilization Plant, which will be used to immobilize up to 50 tonnes of surplus plutonium. The siting for the Plutonium Immobilization Plant will be determined pursuant to the site-specific Surplus Plutonium Disposition Environmental Impact Statement in a Plutonium Deposition Record of Decision in early 1999. This document reflects a new facility using the preferred technology (ceramic immobilization using the can-in-canister approach) and the preferred site (at Savannah River). The Plutonium Immobilization Plant accepts plutonium from pit conversion and from non-pit sources and, through a ceramic immobilization process, converts the plutonium into mineral-like forms that are subsequently encapsulated within a large canister of high-level waste glass. The final immobilized product must make the plutonium as inherently unattractive and inaccessible for use in nuclear weapons as the plutonium in spent fuel from commercial reactors and must be suitable for geologic disposal. Plutonium immobilization at the Savannah River Site uses: (1) A new building, the Plutonium Immobilization Plant, which will convert non-pit surplus plutonium to an oxide form suitable for the immobilization process, immobilize plutonium in a titanate-based ceramic form, place cans of the plutonium-ceramic forms into magazines, and load the magazines into a canister; (2) The existing Defense Waste Processing Facility for the pouring of high-level waste glass into the canisters; and (3) The Actinide Packaging and Storage Facility to receive and store feed materials. The Plutonium Immobilization Plant uses existing Savannah River Site infra-structure for analytical laboratory services, waste handling, fire protection, training, and other support utilities and services. The Plutonium Immobilization Plant

  18. Bimodal, Low Power Pellet Bed Reactor System Design Concept

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.; Liscum-Powell, Jennifer; Pelaccio, Dennis G.

    1994-07-01

    A conceptual design is presented of a bimodal system that employs a pellet bed reactor heat source, helium-xenon Closed Brayton Cycle (CBC) engines, UC fuel, super-alloy structure materials, and hydrogen for propulsion operation. In addition to incorporating state-of-the-art, low risk technologies, and as much off-the-shelf hardware as possible in order to meet a near-term flight demonstration date, the system offers unique design and safety features. These design features include: (a) modularity to support a wide range of electric power and thermal propulsion requirements, (b) sectored, annular reactor core and multiple CBC engines for redundancy and to eliminate a single point failure in the coolant loop, (c) efficient CBC engines, (d) low maximum fuel temperature (<1600 K) that is maintained almost constant during power and propulsion modes, (e) spherical fuel mini-spheres or pellets that provide full retention of fission products and scalability to higher power levels, (f) two independent reactor control systems with built-in redundancy, (h) passive decay heat removal from the reactor core, (g) ground testing of the fully assembled system using electric heaters and unfueled mini-spheres or pellets, (h) negative temperature reactivity feedback for improved reactor operation and safety, (i) high specific impulse (650s-750s) and specific power (11.0- 21.9 We/kg), at relatively low power levels (10-40 kWe).

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

    SciTech Connect

    Hallbert, B.P.; Sebok, A.; Haugset, K.

    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.

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

    SciTech Connect

    Ingersoll, D.T.

    2004-07-29

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

  1. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-01-20

    ... COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.'' DATES... developed using this Catalog along with the Operator Licensing Examination Standards for Power...

  2. REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT

    DOEpatents

    Loeb, E.

    1961-01-17

    A power generator in the form of a flywheel with four reactors positioned about its rim is described. The reactors are so positioned that steam, produced in the reactor, exists tangentially to the flywheel, giving it a rotation. The reactors are incompletely moderated without water. The water enters the flywheel at its axis, under sufficient pressure to force it through the reactors, where it is converted to steam. The fuel consists of parallel twisted ribbons assembled to approximate a cylinder.

  3. Constructal method to optimize solar thermochemical reactor design

    SciTech Connect

    Tescari, S.; Mazet, N.; Neveu, P.

    2010-09-15

    The objective of this study is the geometrical optimization of a thermochemical reactor, which works simultaneously as solar collector and reactor. The heat (concentrated solar radiation) is supplied on a small peripheral surface and has to be dispersed in the entire reactive volume in order to activate the reaction all over the material. A similarity between this study and the point to volume problem analyzed by the constructal approach (Bejan, 2000) is evident. This approach was successfully applied to several domains, for example for the coupled mass and conductive heat transfer (Azoumah et al., 2004). Focusing on solar reactors, this work aims to apply constructal analysis to coupled conductive and radiative heat transfer. As a first step, the chemical reaction is represented by a uniform heat sink inside the material. The objective is to optimize the reactor geometry in order to maximize its efficiency. By using some hypothesis, a simplified solution is found. A parametric study provides the influence of different technical and operating parameters on the maximal efficiency and on the optimal shape. Different reactor designs (filled cylinder, cavity and honeycomb reactors) are compared, in order to determine the most efficient structure according to the operating conditions. Finally, these results are compared with a CFD model in order to validate the assumptions. (author)

  4. Cogeneration of Electricity and Potable Water Using The International Reactor Innovative And Secure (IRIS) Design

    SciTech Connect

    Ingersoll, D.T.; Binder, J.L.; Kostin, V.I.; Panov, Y.K.; Polunichev, V.; Ricotti, M.E.; Conti, D.; Alonso, G.

    2004-10-06

    The worldwide demand for potable water has been steadily growing and is projected to accelerate, driven by a continued population growth and industrialization of emerging countries. This growth is reflected in a recent market survey by the World Resources Institute, which shows a doubling in the installed capacity of seawater desalination plants every ten years. The production of desalinated water is energy intensive, requiring approximately 3-6 kWh/m3 of produced desalted water. At current U.S. water use rates, a dedicated 1000 MW power plant for every one million people would be required to meet our water needs with desalted water. Nuclear energy plants are attractive for large scale desalination application. The thermal energy produced in a nuclear plant can provide both electricity and desalted water without the production of greenhouse gases. A particularly attractive option for nuclear desalination is to couple a desalination plant with an advanced, modular, passively safe reactor design. The use of small-to-medium sized nuclear power plants allows for countries with smaller electrical grid needs and infrastructure to add new electrical and water capacity in more appropriate increments and allows countries to consider siting plants at a broader number of distributed locations. To meet these needs, a modified version of the International Reactor Innovative and Secure (IRIS) nuclear power plant design has been developed for the cogeneration of electricity and desalted water. The modular, passively safe features of IRIS make it especially well adapted for this application. Furthermore, several design features of the IRIS reactor will ensure a safe and reliable source of energy and water even for countries with limited nuclear power experience and infrastructure. The IRIS-D design utilizes low-quality steam extracted from the low-pressure turbine to boil seawater in a multi-effect distillation desalination plant. The desalination plant is based on the horizontal

  5. Fabrication of reactor shields at the Y-12 plant

    NASA Astrophysics Data System (ADS)

    Asbury, W. L.; Kosinski, F. E.; Royer, L. T.

    The Oak Ridge Y-12 Plant is the lead DOE facility for the large scale synthesis, manufacture, and fabrication of lithium hydride/deuteride (LiH/D) materials. Fabrication of large crack free LiH/D parts of uniform density and strength is best achieved by isostatic pressing, using powder metallurgy methods. The highly reactive and moisture sensitive nature of LiH/D has required a number of unique facilities and processing methods at Y-12 to process LiH/D in a safe manner and to maintain high quality products. The history of the fabrication of neutron shields for the SNAP 10A reactor is discussed. Powder metallurgical techniques were used to fabricate that shielding system at Y-12, and the units were tested for shielding efficiency at the Tower Shielding Facility at ORNL.

  6. Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency

    SciTech Connect

    R. Wigeland; K. Hamman

    2009-09-01

    Suggested for Track 7: Advances in Reactor Core Design and In-Core Management _____________________________________________________________________________________ Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency R. Wigeland and K. Hamman Idaho National Laboratory Given the ability of fast reactors to effectively transmute the transuranic elements as are present in spent nuclear fuel, fast reactors are being considered as one element of future nuclear power systems to enable continued use and growth of nuclear power by limiting high-level waste generation. However, a key issue for fast reactors is higher electricity cost relative to other forms of nuclear energy generation. The economics of the fast reactor are affected by the amount of electric power that can be produced from a reactor, i.e., the thermal efficiency for electricity generation. The present study is examining the potential for fast reactor subassembly design changes to improve the thermal efficiency by increasing the average coolant outlet temperature without increasing peak temperatures within the subassembly, i.e., to make better use of current technology. Sodium-cooled fast reactors operate at temperatures far below the coolant boiling point, so that the maximum coolant outlet temperature is limited by the acceptable peak temperatures for the reactor fuel and cladding. Fast reactor fuel subassemblies have historically been constructed using a large number of small diameter fuel pins contained within a tube of hexagonal cross-section, or hexcan. Due to this design, there is a larger coolant flow area next to the hexcan wall as compared to flow area in the interior of the subassembly. This results in a higher flow rate near the hexcan wall, overcooling the fuel pins next to the wall, and a non-uniform coolant temperature distribution. It has been recognized for many years that this difference in sodium coolant temperature was detrimental to achieving

  7. Space Nuclear Power Plant Pre-Conceptual Design Report, For Information

    SciTech Connect

    B. Levine

    2006-01-27

    This letter transmits, for information, the Project Prometheus Space Nuclear Power Plant (SNPP) Pre-Conceptual Design Report completed by the Naval Reactors Prime Contractor Team (NRPCT). This report documents the work pertaining to the Reactor Module, which includes integration of the space nuclear reactor with the reactor radiation shield, energy conversion, and instrumentation and control segments. This document also describes integration of the Reactor Module with the Heat Rejection segment, the Power Conditioning and Distribution subsystem (which comprise the SNPP), and the remainder of the Prometheus spaceship.

  8. Applicability of GALE-86 Codes to Integral Pressurized Water Reactor designs

    SciTech Connect

    Geelhood, Kenneth J.; Rishel, Jeremy P.

    2012-06-01

    This report describes work that Pacific Northwest National Laboratory is doing to assist the U.S. Nuclear Regulatory Commission (NRC) Office of New Reactors (NRO) staff in their reviews of applications for nuclear power plants using new reactor core designs. These designs include small integral PWRs (IRIS, mPower, and NuScale reactor designs), HTGRs, (pebble-bed and prismatic-block modular reactor designs) and SFRs (4S and PRISM reactor designs). Under this specific task, PNNL will assist the NRC staff in reviewing the current versions of the GALE codes and identify features and limitations that would need to be modified to accommodate the technical review of iPWR and mPower® license applications and recommend specific changes to the code, NUREG-0017, and associated NRC guidance. This contract is necessary to support the licensing of iPWRs with a near-term focus on the B&W mPower® reactor design. While the focus of this review is on the mPower® reactor design, the review of the code and the scope of recommended changes consider a revision of the GALE codes that would make them universally applicable for other types of integral PWR designs. The results of a detailed comparison between PWR and iPWR designs are reported here. Also included is an investigation of the GALE code and its basis and a determination as to the applicability of each of the bases to an iPWR design. The issues investigated come from a list provided by NRC staff, the results of comparing the PWR and iPWR designs, the parameters identified as having a large impact on the code outputs from a recent sensitivity study and the main bases identified in NUREG-0017. This report will provide a summary of the gaps in the GALE codes as they relate to iPWR designs and for each gap will propose what work could be performed to fill that gap and create a version of GALE that is applicable to integral PWR designs.

  9. Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

    SciTech Connect

    Qualls, A. L.; Brown, Nicholas R.; Betzler, Benjamin R.; Carbajo, Juan; Hale, Richard Edward; Harrison, Thomas J.; Powers, Jeffrey J.; Robb, Kevin R.; Terrell, Jerry W.; Wysocki, Aaron J.

    2016-02-01

    concept, and it will demonstrate key operational features of that design. The FHR DR will be closely scaled to the SmAHTR concept in power and flows, so any technologies demonstrated will be directly applicable to a reactor concept of that size. The FHR DR is not a commercial prototype design, but rather a DR that serves a cost and risk mitigation function for a later commercial prototype. It is expected to have a limited operational lifetime compared to a commercial plant. It is designed to be a low-cost reactor compared to more mature advanced prototype DRs. A primary reason to build the FHR DR is to learn about salt reactor technologies and demonstrate solutions to remaining technical gaps.

  10. Automated Design and Optimization of Pebble-bed Reactor Cores

    SciTech Connect

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2010-07-01

    We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.

  11. Portfolio Assessment on Chemical Reactor Analysis and Process Design Courses

    ERIC Educational Resources Information Center

    Alha, Katariina

    2004-01-01

    Assessment determines what students regard as important: if a teacher wants to change students' learning, he/she should change the methods of assessment. This article describes the use of portfolio assessment on five courses dealing with chemical reactor and process design during the years 1999-2001. Although the use of portfolio was a new…

  12. Structural Design Challenges in Design Certification Applications for New Reactors

    SciTech Connect

    Miranda, M.; Braverman, J.; Wei, X.; Hofmayer, C.; Xu, J.

    2011-07-17

    The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early reso- lution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific de- sign parameters are confined within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of structural design chal- lenges encountered in recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.

  13. Designing visual displays and system models for safe reactor operations

    SciTech Connect

    Brown-VanHoozer, S.A.

    1995-12-31

    The material presented in this paper is based on two studies involving the design of visual displays and the user`s prospective model of a system. The studies involve a methodology known as Neuro-Linguistic Programming and its use in expanding design choices from the operator`s perspective image. The contents of this paper focuses on the studies and how they are applicable to the safety of operating reactors.

  14. The Virtual Environment for Reactor Applications (VERA). Design and architecture☆

    NASA Astrophysics Data System (ADS)

    Turner, John A.; Clarno, Kevin; Sieger, Matt; Bartlett, Roscoe; Collins, Benjamin; Pawlowski, Roger; Schmidt, Rodney; Summers, Randall

    2016-12-01

    VERA, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors (CASL). CASL was established for the modeling and simulation of commercial nuclear reactors. VERA consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. VERA also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of VERA from both software and numerical perspectives, along with the goals and constraints that drove major design decisions, and their implications. We explain why VERA is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the use of VERA tools for a variety of challenging applications within the nuclear industry.

  15. Design and testing of integrated circuits for reactor protection channels

    SciTech Connect

    Battle, R.E.; Vandermolen, R.I.; Jagadish, U.; Swail, B.K.; Naser, J.; Rana, I.

    1995-06-01

    Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. Purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing.

  16. Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

    SciTech Connect

    1995-08-01

    This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

  17. Mechanical design of a light water breeder reactor

    DOEpatents

    Fauth, Jr., William L.; Jones, Daniel S.; Kolsun, George J.; Erbes, John G.; Brennan, John J.; Weissburg, James A.; Sharbaugh, John E.

    1976-01-01

    In a light water reactor system using the thorium-232 -- uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements.

  18. KEY DESIGN REQUIREMENTS FOR THE HIGH TEMPERATURE GAS-COOLED REACTOR NUCLEAR HEAT SUPPLY SYSTEM

    SciTech Connect

    L.E. Demick

    2010-09-01

    Key requirements that affect the design of the high temperature gas-cooled reactor nuclear heat supply system (HTGR-NHSS) as the NGNP Project progresses through the design, licensing, construction and testing of the first of a kind HTGR based plant are summarized. These requirements derive from pre-conceptual design development completed to-date by HTGR Suppliers, collaboration with potential end users of the HTGR technology to identify energy needs, evaluation of integration of the HTGR technology with industrial processes and recommendations of the NGNP Project Senior Advisory Group.

  19. Modeling and simulation of ammonia removal from purge gases of ammonia plants using a catalytic Pd-Ag membrane reactor.

    PubMed

    Rahimpour, M R; Asgari, A

    2008-05-01

    In this work, the removal of ammonia from synthesis purge gas of an ammonia plant has been investigated. Since the ammonia decomposition is thermodynamically limited, a membrane reactor is used for complete decomposition. A double pipe catalytic membrane reactor is used to remove ammonia from purge gas. The purge gas is flowing in the reaction side and is converted to hydrogen and nitrogen over nickel-alumina catalyst. The hydrogen is transferred through the Pd-Ag membrane of tube side to the shell side. A mathematical model including conservation of mass in the tube and shell side of reactor is proposed. The proposed model was solved numerically and the effects of different parameters on the rector performance were investigated. The effects of pressure, temperature, flow rate (sweep ratio), membrane thickness and reactor diameter have been investigated in the present study. Increasing ammonia conversion was observed by raising the temperature, sweep ratio and reducing membrane thickness. When the pressure increases, the decomposition is gone toward completion but, at low pressure the ammonia conversion in the outset of reactor is higher than other pressures, but complete destruction of the ammonia cannot be achieved. The proposed model can be used for design of an industrial catalytic membrane reactor for removal of ammonia from ammonia plant and reducing NO(x) emissions.

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

    SciTech Connect

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

    1994-06-01

    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.

  1. Winding insulation in electromagnetic systems for Tokamak reactor plants

    NASA Astrophysics Data System (ADS)

    Maslov, V. V.; Trubachev, S. G.

    1985-01-01

    Magnetic containment of the plasma in nuclear fusion reactors of the Tokamak type requires electromagnets with insulation which must withstand high temperatures and thermal shocks as well as ionizing radiation in various forms and electric fields, and mechanical loads. Insulation materials to ensure adequate thermophysical and mechanical properties are evaluated, followed by design of insulation systems with satisfactory performance characteristics. Data on neutron fluence energy characteristics and radiation absorption doses during neutron interactions are essential for such an evaluation. Materials considered for insulation in electromagnets with superconductor and cryoresistance windings are glass mica tape with epoxy compound impregnation, glass cloth with epoxy compound impregnation (STE), polyimide glass cloth with adhesive coating (LSNL), glass Textolite with epoxy phenolic binder (STEN), epoxy resin paste with mineral fillers (PE), and polyurethane compound modified by epoxy resin with mineral filler (KPU).

  2. Modeling the high-temperature gas-cooled reactor process heat plant: a nuclear to chemical conversion process

    SciTech Connect

    Pfremmer, R.D.; Openshaw, F.L.

    1982-05-01

    The high-temperature heat available from the High-Temperature Gas-Cooled Reactor (HTGR) makes it suitable for many process applications. One of these applications is a large-scale energy production plant where nuclear energy is converted into chemical energy and stored for industrial or utility applications. This concept combines presently available nuclear HTGR technology and energy conversion chemical technology. The design of this complex plant involves questions of interacting plant dynamics and overall plant control. This paper discusses how these questions were answered with the aid of a hybrid computer model that was developed within the time-frame of the conceptual design studies. A brief discussion is given of the generally good operability shown for the plant and of the specific potential problems and their anticipated solution. The paper stresses the advantages of providing this information in the earliest conceptual phases of the design.

  3. Modeling the high-temperature gas-cooled reactor process heat plant a nuclear to chemical conversion process

    SciTech Connect

    Pfremmer, R.D.; Openshaw, F.L.

    1982-08-01

    The high-temperature heat available from the high-temperature gas-cooled reactor (HTGR) makes it suitable for many process applications. One of these applications is a large-scale energy production plant where nuclear energy is converted into chemical energy and stored for industrial or utility applications. This concept combines presently available nuclear HTGR technology and energy conversion chemical technology. The design of this complex plant involves questions of interacting plant dynamics and overall plant control. This paper discusses how these questions were answered with the aid of a hybrid computer model that was developed within the time-frame of the conceptual design studies. A brief discussion is given of the generally good operability shown for the plant and of the specific potential problems and their anticipated solution. The paper stresses the advantages of providing this information in the earliest conceptual phases of the design.

  4. Yannawa wastewater treatment plant (Bangkok, Thailand): design, construction and operation.

    PubMed

    Kirkwood, S

    2004-01-01

    Yannawa Wastewater Treatment plant (Phase 1) serves a population equivalent of 500,000 and is located on a restricted site within the city of Bangkok, Thailand. Secondary treatment is based on the CASS sequencing batch reactor (SBR) process and the plant is one of the largest multi-storey SBRs in the world. The limitation of available site area, the ground conditions and the characteristics of the wastewater to be treated set a series of challenges for the designers, contractors and commissioning and operational staff. This paper briefly describes the collection system, the process selection and the treatment streams of the wastewater treatment plant. The SBR secondary treatment plant is described in more detail. The problems that arose during commissioning and operation and the solutions made possible by the use of an SBR type of process are discussed. Details of plant performance during performance testing and during the first three years of plant operation are provided.

  5. Optimized Battery-Type Reactor Primary System Design Utilizing Lead

    SciTech Connect

    Yu, Yong H.; Son, Hyoung M.; Lee, Il S.; Suh, Kune Y.

    2006-07-01

    A number of small and medium size reactors are being developed worldwide as well as large electricity generation reactors for co-generation, district heating or desalination. The Seoul National University has started to develop 23 MWth BORIS (Battery Optimized Reactor Integral System) as a multi-purpose reactor. BORIS is an integral-type optimized fast reactor with an ultra long life core. BORIS is being designed to meet the Generation IV nuclear energy system goals of sustainability, safety, reliability and economics. Major features of BORIS include 20 consecutive years of operation without refueling; elimination of an intermediate heat transport loop and main coolant pump; open core without individual subassemblies; inherent negative reactivity feedback; and inherent load following capability. Its one mission is to provide incremental electricity generation to match the needs of developing nations and especially remote communities without major electrical grid connections. BORIS consists of a reactor module, heat exchanger, coolant module, guard vessel, reactor vessel auxiliary cooling system (RVACS), secondary system, containment and the seismic isolation. BORIS is designed to generate 10 MWe with the resulting thermal efficiency of 45 %. BORIS uses lead as the primary system coolant because of the inherent safety of the material. BORIS is coupled with a supercritical carbon dioxide Brayton cycle as the secondary system to gain a high cycle efficiency in the range of 45 %. The reference core consists of 757 fuel rods without assembly with an active core height of 0.8 m. The BORIS core consists of single enrichment zone composed of a Pu-MA (minor actinides)-U-N fuel and a ferritic-martensitic stainless steel clad. This study is intended to set up appropriate reactor vessel geometry by performing thermal hydraulic analysis on RVACS using computational fluid dynamics codes; to examine the liquid metal coolant behavior along the subchannels; to find out whether the

  6. HYLIFE-II inertial confinement fusion reactor design

    NASA Astrophysics Data System (ADS)

    Moir, R. W.

    1990-12-01

    The HYLIFE-2 inertial fusion power plant design study uses a liquid fall, in the form of jets to protect the first structural wall from neutron damage, x rays, and blast to provide a 30-y lifetime. HYLIFE-1 used liquid lithium. HYLIFE 2 avoids the fire hazard of lithium by using a molten salt composed of fluorine, lithium, and beryllium (Li2, BeF4) called Flibe. Access for heavy-ion beams is provided. Calculations for assumed heavy-ion beam performance show a nominal gain of 70 at 5 MJ producing 350 MJ, about 5.2 times less yield than the 1.8 GJ from a driver energy of 4.5 MJ with gain of 400 for HYLIFE-1. The nominal 1 GWe of power can be maintained by increasing the repetition rate by a factor of about 5.2, from 1.5 to 8 Hz. A higher repetition rate requires faster re-establishment of the jets after a shot, which can be accomplished in part by decreasing the jet fall height and increasing the jet flow velocity. Multiple chambers may be required. In addition, although not considered for HYLIFE-1, there is undoubtedly liquid splash that must be forcibly cleared because gravity is too slow, especially at high repetition rates. Splash removal can be accomplished by either pulsed or oscillating jet flows. The cost of electricity is estimated to be 0.09 $/kW times h in constant 1988 dollars, about twice that of future coal and light water reactor nuclear power. The driver beam cost is about one-half the total cost.

  7. Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

    NASA Astrophysics Data System (ADS)

    Nur Krisna, Dwita; Su’ud, Zaki

    2017-01-01

    Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

  8. REACTOR PRESSURE VESSEL TEMPERATURE ANALYSIS OF CANDIDATE VERY HIGH TEMPERATURE REACTOR DESIGNS

    SciTech Connect

    Hans D. Gougar; Cliff B. Davis; George Hayner; Kevan Weaver

    2006-10-01

    Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code. Because PEBBED-THERMIX has not been extensively validated, confirmatory calculations were also performed with RELAP5-3D for the pebble-bed design. During normal operation, the predicted axial profiles in reactor vessel temperature were similar with both codes and the predicted maximum values were within 2 °C. The trends of the calculated vessel temperatures were similar during the depressurized conduction cooldown accident. The maximum value predicted with RELAP5-3D during the depressurized conduction cooldown accident was about 40 °C higher than that predicted with PEBBED. This agreement is considered reasonable based on the expected uncertainty in either calculation. The differences between the PEBBED and RELAP5-3D calculations were not large enough to affect conclusions concerning comparisons between calculated and allowed maximum temperatures during normal operation and the depressurized conduction cooldown accident.

  9. Nuclear Design of the HOMER-15 Mars Surface Fission Reactor

    SciTech Connect

    Poston, David I.

    2002-07-01

    The next generation of robotic missions to Mars will most likely require robust power sources in the range of 3 to 20 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The goal of this study is to design a compact, low-mass fission system that meets Mars surface power requirements, while maintaining a high level of safety and reliability at a relatively low cost. The Heat pipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive space fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed - which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Heat pipe-Operated Mars Exploration Reactor (HOMER) is a derivative of the HPS designed especially for producing power on the surface of Mars. The HOMER-15 is a 15-kWt reactor that couples with a 3-kWe Stirling engine power system. The reactor contains stainless-steel (SS)-clad uranium nitride (UN) fuel pins that are structurally and thermally bonded to SS/sodium heat pipes. Fission energy is conducted from the fuel pins to the heat pipes, which then carry the heat to the Stirling engine. This paper describes conceptual design and nuclear performance the HOMER-15 reactor. (author)

  10. Preliminary design concept of a subcritical reactor using available resources

    SciTech Connect

    Churnetski, E.L.; Hoyny, V.; Chaudhuri, B.R.; Taprantzis, A.; Yavas, A.

    1993-12-31

    During the Fall 1993 semester, a project was initiated within the Nuclear Engineering Department of the University of Tennessee with the objective of developing a design for a subcritical reactor with maximized multiplication factor using materials currently available. Such a device, if constructed, would serve as a teaching tool for the Department of Nuclear Engineering. Design work was conducted as a large number of computer calculations, with trial pile configurations based on fundamental nuclear engineering principles, in an effort to maximize multiplication factor through fuel element geometry, moderator type, fissile/moderator ratio, and reflector character. The principal objective of the design group for the early phase of this project was to present several possible ``baseline`` reactor designs and identify directions for improvements. For the sake of calculational ease, the cores analyzes to date have been of nearly cubic shape. The SCALE CSAS25 software which runs KENO.Va, a Monte Carlo code, was used for all neutronics calculations. The baseline reactors resulting from work to date are cuboidal in shape and graphite reflected. Two types of fuel element geometries are proposed, a typical triangular pitch rod lattice and an arrangement of discrete fuel slugs placed in a lattice corresponding to body centered cubic packing. The latter arrangement provides slightly higher multiplication factors than the former. Calculations were performed for both graphite and heavy water moderation with heavy water moderation producing considerably higher multiplication factors, as expected. In general, the maximum k{sub eff} for the reactors are in the range of 0.5 to 0.9, well subcritical, except in the cases of the extreme possible values of fuel assay where critical configurations are possible. In these cases, designs with reduced fuel loading are recommended to assure a subcritical multiplication factor.

  11. Thin blanket designs for the Elongated Tokamak Commercial Reactor

    SciTech Connect

    Wong, C.P.C.; Bourque, R.F.; Cheng, E.T.; Creedon, R.L.; Schultz, K.R.

    1986-11-01

    The Elongated Tokamak (ET)* is an innovative concept that uses as highly elongated plasma (plasma height-to-width ratio of 6-10) to allow high plasma current and high toroidal betas. ET has the potential for the development of small-size, high-power density, low-cost fusion reactors using normal conducting coils. The elongated plasma shape is achieved by use of a continuous stack of PF coils parallel to the plasma surface on both inbound and outbound sides. To achieve plasma stability, these coil stacks must be located no further than one plasma minor radius from the plasma edge, greatly restricting the space available for blankets. In order to assess the potential of a small reactor, the authors evaluated and designed blankets 30 to 40 cm thick. Three different thin blanket designs were found to be acceptable: FLiBe self-cooled, helium-cooled lithium, and helium-cooled 17Li83Pb blanket designs. A lithium-cooled integrated blanket-coil design (BLITZ-coil) was also found to be suitable for the ET commercial reactors.

  12. The near boiling reactor: Conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

    NASA Astrophysics Data System (ADS)

    Cole, Christopher J. P.

    Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the Victoria Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96°C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional areas

  13. Upgrading of an activated sludge wastewater treatment plant by adding a moving bed biofilm reactor as pre-treatment and ozonation followed by biofiltration for enhanced COD reduction: design and operation experience.

    PubMed

    Kaindl, Nikolaus

    2010-01-01

    A paper mill producing 500,000 ton of graphic paper annually has an on-site wastewater treatment plant that treats 7,240,000 m³ of wastewater per year, mechanically first, then biologically and at last by ozonation. Increased paper production capacity led to higher COD load in the mill effluent while production of higher proportions of brighter products gave worse biodegradability. Therefore the biological capacity of the WWTP needed to be increased and extra measures were necessary to enhance the efficiency of COD reduction. The full scale implementation of one MBBR with a volume of 1,230 m³ was accomplished in 2000 followed by another MBBR of 2,475 m³ in 2002. An ozonation step with a capacity of 75 kg O₃/h was added in 2004 to meet higher COD reduction demands during the production of brighter products and thus keeping the given outflow limits. Adding a moving bed biofilm reactor prior to the existing activated sludge step gives: (i) cost advantages when increasing biological capacity as higher COD volume loads of MBBRs allow smaller reactors than usual for activated sludge plants; (ii) a relief of strain from the activated sludge step by biological degradation in the MBBR; (iii) equalizing of peaks in the COD load and toxic effects before affecting the activated sludge step; (iv) a stable volume sludge index below 100 ml/g in combination with an optimization of the activated sludge step allows good sludge separation--an important condition for further treatment with ozone. Ozonation and subsequent bio-filtration pre-treated waste water provide: (i) reduction of hard COD unobtainable by conventional treatment; (ii) controllable COD reduction in a very wide range and therefore elimination of COD-peaks; (iii) reduction of treatment costs by combination of ozonation and subsequent bio-filtration; (iv) decrease of the color in the ozonated wastewater. The MBBR step proved very simple to operate as part of the biological treatment. Excellent control of the COD

  14. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    SciTech Connect

    J. M. Beck; L. F. Pincock

    2011-04-01

    The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

  15. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    SciTech Connect

    Qualls, A. L.; Betzler, Benjamin R.; Brown, Nicholas R.; Carbajo, Juan; Greenwood, Michael Scott; Hale, Richard Edward; Harrison, Thomas J.; Powers, Jeffrey J.; Robb, Kevin R.; Terrell, Jerry W.

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  16. Preconceptual design of the new production reactor circulator test facility

    SciTech Connect

    Thurston, G.

    1990-06-01

    This report presents the results of a study of a new circulator test facility for the New Production Reactor Modular High-Temperature Gas-Cooled Reactor. The report addresses the preconceptual design of a stand-alone test facility with all the required equipment to test the Main Circulator/shutoff valve and Shutdown Cooling Circulator/shutoff valve. Each type of circulator will be tested in its own full flow, full power helium test loop. Testing will cover the entire operating range of each unit. The loop will include a test vessel, in which the circulator/valve will be mounted, and external piping. The external flow piping will include a throttle valve, flowmeter, and heat exchanger. Subsystems will include helium handling, helium purification, and cooling water. A computer-based data acquisition and control system will be provided. The estimated costs for the design and construction of this facility are included. 2 refs., 15 figs.

  17. The design and performance of the research reactor fuel counter

    SciTech Connect

    Abhold, M.E.; Hsue, S.T.; Menlove, H.O.; Walton, G.; Holt, S.

    1996-09-01

    This paper describes the design features, hardware specifications, and performance characteristics of the Research Reactor Fuel Counter (RRFC) System. The system is an active mode neutron coincidence counter intended to assay material test reactor fuel assemblies under water. The RRFC contains 12 {sup 3}He tubes, each with its own preamplifier, and a single ion chamber. The neutron counting electronics are based on the Los Alamos Portable Shift Register (PSR) and the gamma readout is a manual-range pico-ammeter of Los Alamos design. The RRFC is connected to the surface by a 20-m-long cable bundle. The PSR is controlled by a portable IBM computer running a modified version of the Los Alamos neutron coincidence counting code also called RRFC. There is a manual that describes the RRFC software.

  18. Non-linear analysis in Light Water Reactor design

    SciTech Connect

    Rashid, Y.R.; Sharabi, M.N.; Nickell, R.E.; Esztergar, E.P.; Jones, J.W.

    1980-03-01

    The results obtained from a scoping study sponsored by the US Department of Energy (DOE) under the Light Water Reactor (LWR) Safety Technology Program at Sandia National Laboratories are presented. Basically, this project calls for the examination of the hypothesis that the use of nonlinear analysis methods in the design of LWR systems and components of interest include such items as: the reactor vessel, vessel internals, nozzles and penetrations, component support structures, and containment structures. Piping systems are excluded because they are being addressed by a separate study. Essentially, the findings were that nonlinear analysis methods are beneficial to LWR design from a technical point of view. However, the costs needed to implement these methods are the roadblock to readily adopting them. In this sense, a cost-benefit type of analysis must be made on the various topics identified by these studies and priorities must be established. This document is the complete report by ANATECH International Corporation.

  19. The design of asymmetric 4 pi shields for space reactors

    NASA Technical Reports Server (NTRS)

    Engle, W. W., Jr.; Childs, R. L.; Mynatt, F. R.

    1972-01-01

    A one dimensional shield optimization program based on the method of discrete ordinates has been developed and is used to determine material thicknesses used in asymmetric 4 pion shields for space power reactors. The two dimensional discrete ordinates program DOT is used to check the design, and the information generated in the DOT calculation is used as a guide in shaping the shield which may be considered a first step in two dimensional shield optimization.

  20. Experiments on rehabilitation of radioactive metallic waste (RMW) of reactor stainless steels of Siberian chemical plant

    NASA Astrophysics Data System (ADS)

    Kolpakov, G. N.; Zakusilov, V. V.; Demyanenko, N. V.; Mishin, A. S.

    2016-06-01

    Stainless steel pipes, used to cool a reactor plant, have a high cost, and after taking a reactor out of service they must be buried together with other radioactive waste. Therefore, the relevant problem is the rinse of pipes from contamination, followed by returning to operation.

  1. Designing the Perfect Plant: Activities to Investigate Plant Ecology

    ERIC Educational Resources Information Center

    Lehnhoff, Erik; Woolbaugh, Walt; Rew, Lisa

    2008-01-01

    Plant ecology is an important subject that often receives little attention in middle school, as more time during science classes is devoted to plant biology. Therefore, the authors have developed a series of activities, including a card game--Designing the Perfect Plant--to introduce student's to plant ecology and the ecological trade offs…

  2. Conceptual design characteristics of a denatured molten-salt reactor with once-through fueling

    SciTech Connect

    Engel, J.R.; Bauman, H.F.; Dearing, J.F.; Grimes, W.R.; McCoy, H.E.; Rhoades, W.A.

    1980-07-01

    A study was made to examine the conceptual feasibility of a molten-salt power reactor fueled with denatured /sup 235/U and operated with a minimum of chemical processing. Because such a reactor would not have a positive breeding gain, reductions in the fuel conversion ratio were allowed in the design to achieve other potentially favorable characteristics for the reactor. A conceptual core design was developed in which the power density was low enough to allow a 30-year life expectancy of the moderator graphite with a fluence limit of 3 x 10/sup 26/ neutrons/m/sup 2/ (E > 50 keV). This reactor could be made critical with about 3450 kg of 20% enriched /sup 235/U and operated for 30 years with routine additions of denatured /sup 235/U and no chemical processing for removal of fission products. A review of the chemical considerations assoicated with the conceptual fuel cycle indicates that no substantial difficulties would be expected if the soluble fission products and higher actinides were allowed to remain in the fuel salt for the life of the plant.

  3. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Galvez, Cristhian

    2011-12-01

    The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the passive safety cooling system with a dual purpose, to assess the capacity to maintain the core at safe temperatures and to assist the design process of this system to achieve this objective. The analysis requires the use of complex computational tools for simulation and verification using analytical solutions and comparisons with experimental data. This investigation builds upon previous detailed design work for the PB-AHTR components, including the core, reactivity control mechanisms and the intermediate heat exchanger, developed in 2008. In addition the study of this reference plant design employs a wealth of auxiliary information including thermal-hydraulic physical phenomena correlations for multiple geometries and thermophysical properties for the constituents of the plant. Finally, the set of performance requirements and limitations imposed from physical constrains and safety considerations provide with a criteria and metrics for acceptability of the design. The passive safety cooling system concept is turned into a detailed design as a result from this study. A methodology for the design of air-cooled passive safety systems was developed and a transient analysis of the plant, evaluating a scrammed loss of forced cooling event was performed. Furthermore, a design optimization study of the passive safety system and an approach for the validation and verification of the analysis is presented. This study demonstrates that the resulting point design responds properly to the

  4. Design and analysis of the SAFE-400 space fission reactor

    NASA Astrophysics Data System (ADS)

    Poston, David I.; Kapernick, Richard J.; Guffee, Ray M.

    2002-01-01

    Ambitious solar system exploration missions in the near future will require robust power sources in the range of 10 to 200 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The Heatpipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive space fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed-which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Safe Affordable Fission Engine (SAFE) is an HPS reactor designed for producing electricity in space. The SAFE-400 is a 400-kWt reactor that has been designed to couple with a 100-kWe Brayton power system. The SAFE-400 contains 127 identical molybdenum (Mo) modules. A Mo/sodium heatpipe is at the center of each module, surrounded by three Mo tubes that each contain a rhenium-clad uranium-nitride fuel pin. Fission energy is conducted from the fuel pins to the heatpipes, which then carry the heat to a heatpipe-to-gas heat exchanger. This paper describes the design and analysis of the current SAFE-400 reactor design. .

  5. A lithium-cooled reactor - Brayton turboelectric power converter design for 100-kWe class space reactor electric systems

    SciTech Connect

    Anderson, R.V.

    1984-08-01

    The conceptual design of a 100-kWe space reactor electric system to satisfy the design goals of the Tri-Agency SP-100 Program has been completed. The system was selected from an initial field of over 500 potential choices covering a wide range of reactor, power converter, shield, heat transport, and radiator subsystems. The selected system -- a lithium-cooled, UN-fueled, refractory-clad reactor coupled to a redundant pair of 110-kWe (gross) Brayton turboelectric power converters -shows strong promise of not only meeting the SP-100 Program design goals but also of providing for substantial growth in power levels for potential future needs.

  6. Survey of light-water-reactor designs to be offered in the United States

    SciTech Connect

    Spiewak, I.

    1986-03-01

    ORNL has conducted a Nuclear Power Options Viability Study for the Department of Energy. That study is primarily concerned with new technology which could be developed for initial operation in the 2000 to 2010 time frame. Such technology would have to compete not only with coal options but with incrementally improved commercial light-water-reactors. This survey reported here was undertaken to gain an understanding of the nuclear commercial technology likely to be offered in the late 1980s and perhaps beyond. The three US vendors actively marketing NSSSs are each developing a product for the future which they expect to be more reliable, more maintainable, more economical, and safer than the present plants. These are all essentially 3800-MW(t) designs, although all are studying smaller plants. They apparently will be off offered as standard prelicensed designs with much larger scope than earlier NSSS offerings, with the possibility of firm prices. Westinghouse with Mitsubishi Heavy Industries is developing a completely new design (APWR) to be built initially in Japan, hopefully for operation by the mid-1990s. Westinghouse is making a strong effort to have the APWR licensed in the US as a standard plant. Combustion Engineering (C-E) is evaluating potential improvements to the System-80 standard design (CESSAR) that has already received final design approval by the NRC. General Electric (GE), with Hitachi and Toshiba, is developing a new design (ABWR) that incorporates advanced features which have been proven by the worldwide BWR suppliers. The ABWR is to be built initially in Japan, but the design could be adapted to the United States. Westinghouse, C-E, and GE have done some conceptual evaluation of reactors in the 600-MW(e) class. The Westinghouse concept is a two-loop plant intended for factory assembly in a shipyard and delivery to a site by barge. The GE concept is a modification of the ABWR with some additional passive safety features. 16 figs.

  7. Plant growth chamber M design

    NASA Technical Reports Server (NTRS)

    Prince, R. P.; Knott, W. M.

    1986-01-01

    Crop production is just one of the many processes involved in establishing long term survival of man in space. The benefits of integrating higher plants into the overall plan was recognized early by NASA through the Closed Ecological Life Support System (CELSS) program. The first step is to design, construct, and operate a sealed (gas, liquid, and solid) plant growth chamber. A 3.6 m diameter by 6.7 m high closed cylinder (previously used as a hypobaric vessel during the Mercury program) is being modified for this purpose. The chamber is mounted on legs with the central axis vertical. Entrance to the chamber is through an airlock. This chamber will be devoted entirely to higher plant experimentation. Any waste treatment, food processing or product storage studies will be carried on outside of this chamber. Its primary purpose is to provide input and output data on solids, liquids, and gases for single crop species and multiple species production using different nutrient delivery systems.

  8. High-Density Plasma Reactors: Simulations for Design

    NASA Technical Reports Server (NTRS)

    Hash, David B.; Meyyappan, Meyya; Arnold, James O. (Technical Monitor)

    1998-01-01

    The development of improved and more efficient plasma reactors is a costly process for the semiconductor industry. Until five years ago, the Industry made most of its advancements through a trial and error approach. More recently, the role of computational modeling in the design process has increased. Both conventional computational fluid dynamics (CFD) techniques like Navier-Stokes solvers as well as particle simulation methods are used to model plasma reactor flowfields. However, since high-density plasma reactors generally operate at low gas pressures on the order of 1 to 10 mTorr, a particle simulation may be necessary because of the failure of CFD techniques to model rarefaction effects. The direct simulation Monte Carlo method is the most widely accepted and employed particle simulation tool and has previously been used to investigate plasma reactor flowfields. A plasma DSMC code is currently under development at NASA Ames Research Center with its foundation as the object-oriented parallel Cornell DSMC code, MONACO. The present investigation is a follow up of a neutral flow investigation of the effects of process parameters as well as reactor design on etch rate and etch rate uniformity. The previous work concentrated on silicon etch of a chlorine flow in a configuration typical of electron cyclotron resonance (ECR) or helical resonator type reactors. The effects of the plasma on the dissociation chemistry were modeled by making assumptions about the electron temperature and number density. The electrons or ions themselves were not simulated.The present work extends these results by simulating the charged species.The electromagnetic fields are calculated such that power deposition is modeled self-consistently. Electron impact reactions are modeled along with mechanisms for charge exchange. An bipolar diffusion assumption is made whereby electrons remain tied to the ions. However, the velocities of tile electrons are allowed to be modified during collisions

  9. Reactor design considerations in mineral sequestration of carbon dioxide

    SciTech Connect

    Ityokumbul, M.T.; Chander, S.; O'Connor, William K.; Dahlin, David C.; Gerdemann, Stephen J.

    2001-01-01

    One of the promising approaches to lowering the anthropogenic carbon dioxide levels in the atmosphere is mineral sequestration. In this approach, the carbon dioxide reacts with alkaline earth containing silicate minerals forming magnesium and/or calcium carbonates. Mineral carbonation is a multiphase reaction process involving gas, liquid and solid phases. The effective design and scale-up of the slurry reactor for mineral carbonation will require careful delineation of the rate determining step and how it changes with the scale of the reactor. The shrinking core model was used to describe the mineral carbonation reaction. Analysis of laboratory data indicates that the transformations of olivine and serpentine are controlled by chemical reaction and diffusion through an ash layer respectively. Rate parameters for olivine and serpentine carbonation are estimated from the laboratory data.

  10. Conceptual design study of the moderate size superconducting spherical tokamak power plant

    NASA Astrophysics Data System (ADS)

    Gi, Keii; Ono, Yasushi; Nakamura, Makoto; Someya, Youji; Utoh, Hiroyasu; Tobita, Kenji; Ono, Masayuki

    2015-06-01

    A new conceptual design of the superconducting spherical tokamak (ST) power plant was proposed as an attractive choice for tokamak fusion reactors. We reassessed a possibility of the ST as a power plant using the conservative reactor engineering constraints often used for the conventional tokamak reactor design. An extensive parameters scan which covers all ranges of feasible superconducting ST reactors was completed, and five constraints which include already achieved plasma magnetohydrodynamic (MHD) and confinement parameters in ST experiments were established for the purpose of choosing the optimum operation point. Based on comparison with the estimated future energy costs of electricity (COEs) in Japan, cost-effective ST reactors can be designed if their COEs are smaller than 120 mills kW-1 h-1 (2013). We selected the optimized design point: A = 2.0 and Rp = 5.4 m after considering the maintenance scheme and TF ripple. A self-consistent free-boundary MHD equilibrium and poloidal field coil configuration of the ST reactor were designed by modifying the neutral beam injection system and plasma profiles. The MHD stability of the equilibrium was analysed and a ramp-up scenario was considered for ensuring the new ST design. The optimized moderate-size ST power plant conceptual design realizes realistic plasma and fusion engineering parameters keeping its economic competitiveness against existing energy sources in Japan.

  11. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR)...

  12. Simplifying Chemical Reactor Design by using Molar Quantities Instead of Fractional Conversion.

    ERIC Educational Resources Information Center

    Brown, Lee F.; Falconer, John L.

    1987-01-01

    Explains the advantages of using molar quantities in chemical reactor design. Advocates the use of differential versions of reactor mass balances rather than the integrated forms. Provides specific examples and cases to illustrate the principles. (ML)

  13. Tokamak reactor studies

    SciTech Connect

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features.

  14. Tritium pellet injector design for tokamak fusion test reactor

    SciTech Connect

    Fisher, P.W.; Baylor, L.R.; Bryan, W.E.; Combs, S.K.; Easterly, C.E.; Lunsford, R.V.; Milora, S.L.; Schuresko, D.D.; White, J.A.; Williamson, D.H.

    1985-01-01

    A tritium pellet injector (TPI) system has been designed for the Tokamak Fusion Test Reactor (TFTR) Q approx. 1 phase of operation. The injector gun utilizes a radial design with eight independent barrels and a common extruder to minimize tritium inventory. The injection line contains guide tubes with intermediate vacuum pumping stations and fast valves to minimize propellant leakage to the torus. The vacuum system is designed for tritium compatibility. The entire injector system is contained in a glove box for secondary containment protection against tritium release. Failure modes and effects have been analyzed, and structural analysis has been performed for most intense predicted earthquake conditions. Details of the design and operation of this system are presented in this paper.

  15. The Virtual Environment for Reactor Applications (VERA): Design and architecture

    SciTech Connect

    Turner, John A.; Clarno, Kevin; Sieger, Matt; Bartlett, Roscoe; Collins, Benjamin; Pawlowski, Roger; Schmidt, Rodney; Summers, Randall

    2016-09-08

    VERA, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors (CASL), the first DOE Hub, which was established in July 2010 for the modeling and simulation of commercial nuclear reactors. VERA consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. VERA also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of VERA from both a software and a numerical perspective, along with the goals and constraints that drove the major design decisions and their implications. As a result, we explain why VERA is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the application of VERA tools for a variety of challenging problems within the nuclear industry.

  16. The Virtual Environment for Reactor Applications (VERA): Design and architecture

    DOE PAGES

    Turner, John A.; Clarno, Kevin; Sieger, Matt; ...

    2016-09-08

    VERA, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors (CASL), the first DOE Hub, which was established in July 2010 for the modeling and simulation of commercial nuclear reactors. VERA consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. VERA also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of VERA from both amore » software and a numerical perspective, along with the goals and constraints that drove the major design decisions and their implications. As a result, we explain why VERA is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the application of VERA tools for a variety of challenging problems within the nuclear industry.« less

  17. Experimental and design experience with passive safety features of liquid metal reactors

    SciTech Connect

    Lucoff, D.M.; Waltar, A.E.; Sackett, J.I.; Salvatores, M.; Aizawa, K.

    1992-10-01

    Liquid metal cooled reactors (LMRs) have already been demonstrated to be robust machines. Many reactor designers now believe that it is possible to include in this technology sufficient passive safety that LMRs would be able to survive loss of flow, loss of heat sink, and transient overpower events, even if the plant protective system fails completely and do so without damage to the core. Early whole-core testing in Rapsodie, EBR-II. and FFTF indicate such designs may be possible. The operational safety testing program in EBR-II is demonstrating benign response of the reactor to a full range of controls failures. But additional testing is needed if transient core structural response under major accident conditions is to be properly understood. The proposed international Phase IIB passive safety tests in FFTF, being designed with a particular emphasis on providing, data to understand core bowing extremes, and further tests planned in EBR-II with processed IFR fuel should provide a substantial and unique database for validating the computer codes being used to simulate postulated accident conditions.

  18. Lunar Nuclear Power Plant With Solid Core Reactor, Heatpipes and Thermoelectric Conversion

    NASA Astrophysics Data System (ADS)

    Sayre, Edwin D.; Ring, Peter J.; Brown, Neil; Elsner, Norbert B.; Bass, John C.

    2008-01-01

    This is a lunar nuclear power plant with the advantages of minimum mass, with no moving parts, no pumped liquid coolant, a solid metal rugged core, with no single point of failure. The electrical output is 100 kilowatts with a 500 kilowatt thermal reactor. The thermoelectric converters surround the potassium heatpipes from the core and water heatpipes surround the converter and connect to the radiator. The solid core reactor is made from HT9 alloy. The fuel is uranium oxide with 90% enrichment. The thermoelectric converter is bonded to the outside of the 1.10 inch ID heat pipe and is 30 inches long. The thermoelectric couple is Si/SiGe-Si/SiC Quantum Well with over 20% efficiency with an 890 K hot side and a 490 K cold side and produces 625 Watts. 176 converters produce 110 kWe. With less than 10% loss in controls this yields 100 kWe for use. The cylindrical thermoelectric converter is designed and fabricated by HIPing to keep brittle materials in compression and to ensure conductivity. The solid core is fabricated by machining the heatpipe tubes with 6 grooves that are diffusion bonded together by HIPing to form the fuel tubes. The maximum temperature of the heat pipes is 940 K and the return flow temperature is 890 K. The reactor core is hexagonal shaped, 61 cm. wide and 76.2 cm high with 12 rotating control drums surrounding it. There is shielding to protect components and human habitation. The radiator is daisy shaped at 45 degrees with each petal 5.5 meters long. The design life is ten years.

  19. Lunar Nuclear Power Plant With Solid Core Reactor, Heatpipes and Thermoelectric Conversion

    SciTech Connect

    Sayre, Edwin D.; Ring, Peter J.; Brown, Neil; Elsner, Norbert B.; Bass, John C.

    2008-01-21

    This is a lunar nuclear power plant with the advantages of minimum mass, with no moving parts, no pumped liquid coolant, a solid metal rugged core, with no single point of failure. The electrical output is 100 kilowatts with a 500 kilowatt thermal reactor. The thermoelectric converters surround the potassium heatpipes from the core and water heatpipes surround the converter and connect to the radiator. The solid core reactor is made from HT9 alloy. The fuel is uranium oxide with 90% enrichment. The thermoelectric converter is bonded to the outside of the 1.10 inch ID heat pipe and is 30 inches long. The thermoelectric couple is Si/SiGe-Si/SiC Quantum Well with over 20% efficiency with an 890 K hot side and a 490 K cold side and produces 625 Watts. 176 converters produce 110 kWe. With less than 10% loss in controls this yields 100 kWe for use. The cylindrical thermoelectric converter is designed and fabricated by HIPing to keep brittle materials in compression and to ensure conductivity. The solid core is fabricated by machining the heatpipe tubes with 6 grooves that are diffusion bonded together by HIPing to form the fuel tubes. The maximum temperature of the heat pipes is 940 K and the return flow temperature is 890 K. The reactor core is hexagonal shaped, 61 cm. wide and 76.2 cm high with 12 rotating control drums surrounding it. There is shielding to protect components and human habitation. The radiator is daisy shaped at 45 degrees with each petal 5.5 meters long. The design life is ten years.

  20. Design and construction of a 7,500 liter immobilized cell reactor-separator for ethanol production from whey

    SciTech Connect

    Dale, M.C.

    1992-12-31

    A 7,500 liter reactor/separator has been constructed for the production of ethanol from concentrated whey permeate. This unit is sited in Hopkinton IA, across the street from a whey generating cheese plant A two phase construction project consisting of (1) building and testing a reactor/separator with a solvent absorber in a single unified housing, and (2) building and testing an extractive distillation/product stripper for the recovery of anhydrous ethanol is under way. The design capacity of this unit is 250,000 gal/yr of anhydrous product. Design and construction details of the reactor/absorber separator are given, and design parameters for the extractive distillation system are described.

  1. Advanced Neutron Source: Plant Design Requirements. Revision 4

    SciTech Connect

    Not Available

    1990-07-01

    The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS.

  2. Conceptual design study of Fusion Experimental Reactor (FY86 FER): Safety

    NASA Astrophysics Data System (ADS)

    Seki, Yasushi; Iida, Hiromasa; Honda, Tsutomu

    1987-08-01

    This report describes the study on safety for FER (Fusion Experimental Reactor) which has been designed as a next step machine to the JT-60. Though the final purpose of this study is to have an image of design base accident, maximum credible accident and to assess their risk or probability, etc., as FER plant system, the emphasis of this years study is placed on fuel-gas circulation system where the tritium inventory is maximum. The report consists of two chapters. The first chapter summarizes the FER system and describes FMEA (Failure Mode and Effect Analysis) and related accident progression sequence for FER plant system as a whole. The second chapter of this report is focused on fuel-gas circulation system including purification, isotope separation and storage. Probability of risk is assessed by the probabilistic risk analysis (PRA) procedure based on FMEA, ETA and FTA.

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

    SciTech Connect

    Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville; Gougar, Hans David; Strydom, Gerhard

    2016-01-01

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

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

    SciTech Connect

    Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville; Gougar, Hans David; Kinsey, J.; Strydom, Gerhard

    2016-03-01

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

  5. Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

    SciTech Connect

    Lata

    1996-09-26

    This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

  6. 75 FR 38564 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Plant Operations...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-07-02

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Plant Operations and Fire Protection The ACRS Subcommittee on Plant Operations and Fire Protection will hold a...

  7. Individual plant examination program: Perspectives on reactor safety and plant performance. Part 1: Final summary report; Volume 1

    SciTech Connect

    1997-12-01

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC`s overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively.

  8. Individual plant examination program: Perspectives on reactor safety and plant performance. Parts 2--5: Final report; Volume 2

    SciTech Connect

    1997-12-01

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC`s overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively.

  9. 10 CFR Appendix N to Part 50 - Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... FACILITIES Pt. 50, App. N Appendix N to Part 50—Standardization of Nuclear Power Plant Designs: Permits...

  10. 10 CFR Appendix N to Part 50 - Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... FACILITIES Pt. 50, App. N Appendix N to Part 50—Standardization of Nuclear Power Plant Designs: Permits...

  11. 10 CFR Appendix N to Part 50 - Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... FACILITIES Pt. 50, App. N Appendix N to Part 50—Standardization of Nuclear Power Plant Designs: Permits...

  12. 10 CFR Appendix N to Part 50 - Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... FACILITIES Pt. 50, App.N Appendix N to Part 50—Standardization of Nuclear Power Plant Designs: Permits...

  13. Passive Safety Features in Advanced Nuclear Power Plant Design

    NASA Astrophysics Data System (ADS)

    Tahir, M.; Chughtai, I. R.; Aslam, M.

    2013-03-01

    For implementation of advance passive safety features in future nuclear power plant design, a passive safety system has been proposed and its response has been observed for Loss of Coolant Accident (LOCA) in the cold leg of a reactor coolant system. In a transient simulation the performance of proposed system is validated against existing safety injection system for a reference power plant of 325 MWe. The existing safety injection system is a huge system and consists of many active components including pumps, valves, piping and Instrumentation and Control (I&C). A good running of the active components of this system is necessary for its functionality as high head safety injection system under design basis accidents. Using reactor simulation technique, the proposed passive safety injection system and existing safety injection system are simulated and tested for their performance under large break LOCA for the same boundary conditions. Critical thermal hydraulic parameters of both the systems are presented graphically and discussed. The results obtained are approximately the same in both the cases. However, the proposed passive safety injection system is a better choice for such type of reactors due to reduction in components with improved safety.

  14. Design Concept for a Nuclear Reactor-Powered Mars Rover

    NASA Astrophysics Data System (ADS)

    Elliott, John O.; Lipinski, Ronald J.; Poston, David I.

    2003-01-01

    A study was recently carried out by a team from JPL and the DOE to investigate the utility of a DOE-developed 3 kWe surface fission power system for Mars missions. The team was originally tasked to perform a study to evaluate the usefulness and feasibility of incorporation of such a power system into a landed mission. In the course of the study it became clear that the application of such a power system was enabling to a wide variety of potential missions. Of these, two missions were developed, one for a stationary lander and one for a reactor-powered rover. This paper discusses the design of the rover mission, which was developed around the concept of incorporating the fission power system directly into a large rover chassis to provide high power, long range traverse capability. The rover design is based on a minimum extrapolation of technology, and adapts existing concepts developed at JPL for the 2009 Mars Science Laboratory (MSL) rover, lander and EDL systems. The small size of the reactor allowed its incorporation directly into an existing large MSL rover chassis design, allowing direct use of MSL aeroshell and pallet lander elements, beefed up to support the significantly greater mass involved in the nuclear power system and its associated shielding. This paper describes the unique design challenges encountered in the development of this mission architecture and incorporation of the fission power system in the rover, and presents a detailed description of the final design of this innovative concept for providing long range, long duration mobility on Mars.

  15. Use of computational fluid dynamics simulations for design of a pretreatment screw conveyor reactor.

    PubMed

    Berson, R Eric; Hanley, Thomas R

    2005-01-01

    Computational fluid dynamics simulations were employed to compare performance of various designs of a pretreatment screw conveyor reactor. The reactor consisted of a vertical screw used to create cross flow between the upward conveying solids and the downward flow of acid. Simulations were performed with the original screw design and a modified design in which the upper flights of the screw were removed. Results of the simulations show visually that the modified design provided favorable plug flow behavior within the reactor. Pressure drop across the length of the reactor without the upper screws in place was predicted by the simulations to be 5 vs 40 kPa for the original design.

  16. Design optimization of metal getter reactors for removing tritium from flowing gas streams

    SciTech Connect

    Nobile, A.; Bieniewski, T.; Frame, K.; Little, R.; Fisher, K.

    1995-10-01

    A reaction engineering approach was used to design a SAES St 198 metal getter reactor for a glovebox detritiation system. The detritiation system will be used to decontaminate and decommission an Li(D, T)-contaminated glovebox previously used in the U.S. nuclear weapons program. The approach involved development of a model that calculates reactor breakthrough curves as a function of various reactor physical parameters. Experiments involving flow of deuterium in nitrogen through a small metal getter reactor validated the model. The model was then used to investigate the effects of temperature, getter pellet size, reactor diameter, and reactor volume on the reactor performance. The resulting design was a 7 cm diam. by 40 cm long cylindrical reactor that operates at 250 {degree}C, and is filled with 5 kg of as-received SAES St 198 getter pellets. The reactor handles a flow rate of 100 L/min. An St 909 getter reactor was used upstream of the St 198 reactor for impurity removal and water decomposition. The glovebox cleanup system design and getter reactor mechanical design are discussed. 14 refs., 6 figs.

  17. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    SciTech Connect

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  18. Design of a Gas Test Loop Facility for the Advanced Test Reactor

    SciTech Connect

    C. A. Wemple

    2005-09-01

    The Office of Nuclear Energy within the U.S. Department of Energy (DOE-NE) has identified the need for irradiation testing of nuclear fuels and materials, primarily in support of the Generation IV (Gen-IV) and Advanced Fuel Cycle Initiative (AFCI) programs. These fuel development programs require a unique environment to test and qualify potential reactor fuel forms. This environment should combine a high fast neutron flux with a hard neutron spectrum and high irradiation temperature. An effort is presently underway at the Idaho National Laboratory (INL) to modify a large flux trap in the Advanced Test Reactor (ATR) to accommodate such a test facility [1,2]. The Gas Test Loop (GTL) Project Conceptual Design was initiated to determine basic feasibility of designing, constructing, and installing in a host irradiation facility, an experimental vehicle that can replicate with reasonable fidelity the fast-flux test environment needed for fuels and materials irradiation testing for advanced reactor concepts. Such a capability will be needed if programs such as the AFCI, Gen-IV, the Next Generation Nuclear Plant (NGNP), and space nuclear propulsion are to meet development objectives and schedules. These programs are beginning some irradiations now, but many call for fast flux testing within this decade.

  19. Moving bed biofilm reactor technology: process applications, design, and performance.

    PubMed

    McQuarrie, James P; Boltz, Joshua P

    2011-06-01

    The moving bed biofilm reactor (MBBR) can operate as a 2- (anoxic) or 3-(aerobic) phase system with buoyant free-moving plastic biofilm carriers. These systems can be used for municipal and industrial wastewater treatment, aquaculture, potable water denitrification, and, in roughing, secondary, tertiary, and sidestream applications. The system includes a submerged biofilm reactor and liquid-solids separation unit. The MBBR process benefits include the following: (1) capacity to meet treatment objectives similar to activated sludge systems with respect to carbon-oxidation and nitrogen removal, but requires a smaller tank volume than a clarifier-coupled activated sludge system; (2) biomass retention is clarifier-independent and solids loading to the liquid-solids separation unit is reduced significantly when compared with activated sludge systems; (3) the MBBR is a continuous-flow process that does not require a special operational cycle for biofilm thickness, L(F), control (e.g., biologically active filter backwashing); and (4) liquid-solids separation can be achieved with a variety of processes, including conventional and compact high-rate processes. Information related to system design is fragmented and poorly documented. This paper seeks to address this issue by summarizing state-of-the art MBBR design procedures and providing the reader with an overview of some commercially available systems and their components.

  20. Reactor design for minimizing product inhibition during enzymatic lignocellulose hydrolysis: II. Quantification of inhibition and suitability of membrane reactors.

    PubMed

    Andrić, Pavle; Meyer, Anne S; Jensen, Peter A; Dam-Johansen, Kim

    2010-01-01

    Product inhibition of cellulolytic enzymes affects the efficiency of the biocatalytic conversion of lignocellulosic biomass to ethanol and other valuable products. New strategies that focus on reactor designs encompassing product removal, notably glucose removal, during enzymatic cellulose conversion are required for alleviation of glucose product inhibition. Supported by numerous calculations this review assesses the quantitative aspects of glucose product inhibition on enzyme-catalyzed cellulose degradation rates. The significance of glucose product inhibition on dimensioning of different ideal reactor types, i.e. batch, continuous stirred, and plug-flow, is illustrated quantitatively by modeling different extents of cellulose conversion at different reaction conditions. The main operational challenges of membrane reactors for lignocellulose conversion are highlighted. Key membrane reactor features, including system set-up, dilution rate, glucose output profile, and the problem of cellobiose are examined to illustrate the quantitative significance of the glucose product inhibition and the total glucose concentration on the cellulolytic conversion rate. Comprehensive overviews of the available literature data for glucose removal by membranes and for cellulose enzyme stability in membrane reactors are given. The treatise clearly shows that membrane reactors allowing continuous, complete, glucose removal during enzymatic cellulose hydrolysis, can provide for both higher cellulose hydrolysis rates and higher enzyme usage efficiency (kg(product)/kg(enzyme)). Current membrane reactor designs are however not feasible for large scale operations. The report emphasizes that the industrial realization of cellulosic ethanol requires more focus on the operational feasibility within the different hydrolysis reactor designs, notably for membrane reactors, to achieve efficient enzyme-catalyzed cellulose degradation.

  1. Evaluation of the applicability of existing nuclear power plant regulatory requirements in the U.S. to advanced small modular reactors.

    SciTech Connect

    LaChance, Jeffrey L.; Wheeler, Timothy A.; Farnum, Cathy Ottinger; Middleton, Bobby D.; Jordan, Sabina Erteza; Duran, Felicia Angelica; Baum, Gregory A.

    2013-05-01

    The current wave of small modular reactor (SMR) designs all have the goal of reducing the cost of management and operations. By optimizing the system, the goal is to make these power plants safer, cheaper to operate and maintain, and more secure. In particular, the reduction in plant staffing can result in significant cost savings. The introduction of advanced reactor designs and increased use of advanced automation technologies in existing nuclear power plants will likely change the roles, responsibilities, composition, and size of the crews required to control plant operations. Similarly, certain security staffing requirements for traditional operational nuclear power plants may not be appropriate or necessary for SMRs due to the simpler, safer and more automated design characteristics of SMRs. As a first step in a process to identify where regulatory requirements may be met with reduced staffing and therefore lower cost, this report identifies the regulatory requirements and associated guidance utilized in the licensing of existing reactors. The potential applicability of these regulations to advanced SMR designs is identified taking into account the unique features of these types of reactors.

  2. Titer-plate formatted continuous flow thermal reactors: Design and performance of a nanoliter reactor.

    PubMed

    Chen, Pin-Chuan; Park, Daniel S; You, Byoung-Hee; Kim, Namwon; Park, Taehyun; Soper, Steven A; Nikitopoulos, Dimitris E; Murphy, Michael C

    2010-08-06

    Arrays of continuous flow thermal reactors were designed, configured, and fabricated in a 96-device (12 × 8) titer-plate format with overall dimensions of 120 mm × 96 mm, with each reactor confined to a 8 mm × 8 mm footprint. To demonstrate the potential, individual 20-cycle (740 nL) and 25-cycle (990 nL) reactors were used to perform the continuous flow polymerase chain reaction (CFPCR) for amplification of DNA fragments of different lengths. Since thermal isolation of the required temperature zones was essential for optimal biochemical reactions, three finite element models, executed with ANSYS (v. 11.0, Canonsburg, PA), were used to characterize the thermal performance and guide system design: (1) a single device to determine the dimensions of the thermal management structures; (2) a single CFPCR device within an 8 mm × 8 mm area to evaluate the integrity of the thermostatic zones; and (3) a single, straight microchannel representing a single loop of the spiral CFPCR device, accounting for all of the heat transfer modes, to determine whether the PCR cocktail was exposed to the proper temperature cycling. In prior work on larger footprint devices, simple grooves between temperature zones provided sufficient thermal resistance between zones. For the small footprint reactor array, 0.4 mm wide and 1.2 mm high fins were necessary within the groove to cool the PCR cocktail efficiently, with a temperature gradient of 15.8°C/mm, as it flowed from the denaturation zone to the renaturation zone. With temperature tolerance bands of ±2°C defined about the nominal temperatures, more than 72.5% of the microchannel length was located within the desired temperature bands. The residence time of the PCR cocktail in each temperature zone decreased and the transition times between zones increased at higher PCR cocktail flow velocities, leading to less time for the amplification reactions. Experiments demonstrated the performance of the CFPCR devices as a function of flow

  3. Titer-plate formatted continuous flow thermal reactors: Design and performance of a nanoliter reactor

    PubMed Central

    Chen, Pin-Chuan; Park, Daniel S.; You, Byoung-Hee; Kim, Namwon; Park, Taehyun; Soper, Steven A.; Nikitopoulos, Dimitris E.; Murphy, Michael C.

    2010-01-01

    Arrays of continuous flow thermal reactors were designed, configured, and fabricated in a 96-device (12 × 8) titer-plate format with overall dimensions of 120 mm × 96 mm, with each reactor confined to a 8 mm × 8 mm footprint. To demonstrate the potential, individual 20-cycle (740 nL) and 25-cycle (990 nL) reactors were used to perform the continuous flow polymerase chain reaction (CFPCR) for amplification of DNA fragments of different lengths. Since thermal isolation of the required temperature zones was essential for optimal biochemical reactions, three finite element models, executed with ANSYS (v. 11.0, Canonsburg, PA), were used to characterize the thermal performance and guide system design: (1) a single device to determine the dimensions of the thermal management structures; (2) a single CFPCR device within an 8 mm × 8 mm area to evaluate the integrity of the thermostatic zones; and (3) a single, straight microchannel representing a single loop of the spiral CFPCR device, accounting for all of the heat transfer modes, to determine whether the PCR cocktail was exposed to the proper temperature cycling. In prior work on larger footprint devices, simple grooves between temperature zones provided sufficient thermal resistance between zones. For the small footprint reactor array, 0.4 mm wide and 1.2 mm high fins were necessary within the groove to cool the PCR cocktail efficiently, with a temperature gradient of 15.8°C/mm, as it flowed from the denaturation zone to the renaturation zone. With temperature tolerance bands of ±2°C defined about the nominal temperatures, more than 72.5% of the microchannel length was located within the desired temperature bands. The residence time of the PCR cocktail in each temperature zone decreased and the transition times between zones increased at higher PCR cocktail flow velocities, leading to less time for the amplification reactions. Experiments demonstrated the performance of the CFPCR devices as a function of flow

  4. Design of Recycle Pressurized Water Reactor with Heavy Water Moderation

    SciTech Connect

    Hibi, Koki; Uchita, Masato

    2004-03-15

    This study presents the conceptual design of the recycle pressurized water reactor (RPWR), which is an innovative PWR fueled with mixed oxide, moderated by heavy water, and having breeding ratios around 1.1. Most of the systems of RPWR can employ those of PWRs. The RPWR has no boric acid systems and has a small tritium removal system. The construction and operation costs would be similar to those of current PWRs. Heavy water cost has decreased drastically with up-to-date producing methods. The reliability of the systems of the RPWR is high, and the research and development cost for RPWR is very low because the core design is fundamentally based on the current PWR technology.

  5. Component and System Sensitivity Considerations for Design of a Lunar ISRU Oxygen Production Plant

    NASA Technical Reports Server (NTRS)

    Linne, Diane L.; Gokoglu, Suleyman; Hegde, Uday G.; Balasubramaniam, Ramaswamy; Santiago-Maldonado, Edgardo

    2009-01-01

    Component and system sensitivities of some design parameters of ISRU system components are analyzed. The differences between terrestrial and lunar excavation are discussed, and a qualitative comparison of large and small excavators is started. The effect of excavator size on the size of the ISRU plant's regolith hoppers is presented. Optimum operating conditions of both hydrogen and carbothermal reduction reactors are explored using recently developed analytical models. Design parameters such as batch size, conversion fraction, and maximum particle size are considered for a hydrogen reduction reactor while batch size, conversion fraction, number of melt zones, and methane flow rate are considered for a carbothermal reduction reactor. For both reactor types the effect of reactor operation on system energy and regolith delivery requirements is presented.

  6. DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

    SciTech Connect

    Pan, Paul Y

    2010-12-10

    An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

  7. Activation product safety in the ARIES-I reactor design

    SciTech Connect

    Herring, J.S. ); Sze, D.K. ); Wong, C.; Cheng, E.T. ); Grotz, S.P. )

    1990-01-01

    The ARIES design effort has sought to maximize the environmental and safety advantages of fusion through careful selection of materials and careful design. Three goals are that the reactor achieve inherent or passive safety, that no public evacuation plan be necessary and that the waste be disposable as 10CFR61 Class C waste. The ARIES-I reactor consists of a SiC composite structure for the first wall and blanket, cooled by 10 MPa He. The breeder is Li{sub 2}ZrO{sub 3}, although Li{sub 2}O and Li{sub 4}SiO{sub 4} were also considered. The divertor consists of SiC composite tubes coated with 2 mm of tungsten. Due to the minimal afterheat of this blanket design, LOCA calculations indicate maximum temperatures will not cause damage if the plasma is promptly extinguished. Two primary safety issues are the zirconium in the breeder and tungsten on the divertor. Li{sub 2}ZrO{sub 3} was chosen because of its demonstrated high-temperature stability. The other breeders have lower afterheat and activation. Use of zirconium in the breeder will necessitate isotopic tailoring to remove {sup 90}Zr and {sup 94}Zr. The 5.8 tonnes of W on the divertor would also have to be tailored to remove {sup 186}W and/or to concentrate {sup 183}W. Thus the ARIES-I design achieves the passive safety and low-level waste disposal criteria with respect to activation products. Development of low activation materials to replace zirconium and tungsten is needed to avoid requiring an evacuation plan.

  8. Uncertainty in bulk-liquid hydrodynamics and biofilm dynamics creates uncertainties in biofilm reactor design.

    PubMed

    Boltz, J P; Daigger, G T

    2010-01-01

    While biofilm reactors may be classified as one of seven different types, the design of each is unified by fundamental biofilm principles. It follows that state-of-the art design of each biofilm reactor type is subject to the same uncertainties (although the degree of uncertainty may vary). This paper describes unifying biofilm principles and uncertainties of importance in biofilm reactor design. This approach to biofilm reactor design represents a shift from the historical approach which was based on empirical criteria and design formulations. The use of such design criteria was largely due to inherent uncertainty over reactor-scale hydrodynamics and biofilm dynamics, which correlate with biofilm thickness, structure and function. An understanding of two fundamental concepts is required to rationally design biofilm reactors: bioreactor hydrodynamics and biofilm dynamics (with particular emphasis on mass transfer resistances). Bulk-liquid hydrodynamics influences biofilm thickness control, surface area, and development. Biofilm dynamics influences biofilm thickness, structure and function. While the complex hydrodynamics of some biofilm reactors such as trickling filters and biological filters have prevented the widespread use of fundamental biofilm principles and mechanistic models in practice, reactors utilizing integrated fixed-film activated sludge or moving bed technology provide a bulk-liquid hydrodynamic environment allowing for their application. From a substrate transformation perspective, mass transfer in biofilm reactors defines the primary difference between suspended growth and biofilm systems: suspended growth systems are kinetically (i.e., biomass) limited and biofilm reactors are primarily diffusion (i.e., biofilm growth surface area) limited.

  9. 10 CFR Appendix A to Part 50 - General Design Criteria for Nuclear Power Plants

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... Containment Atmosphere Cleanup Systems 42 Testing of Containment Atmosphere Cleanup Systems 43 Cooling Water 44 Inspection of Cooling Water System 45 Testing of Cooling Water System 46 V. Reactor Containment... requirements for the principal design criteria for water-cooled nuclear power plants similar in design...

  10. Conceptual Design of Passive Safety System for Lead-Bismuth Cooled Fast Reactor

    NASA Astrophysics Data System (ADS)

    Abdullah, A. G.; Nandiyanto, A. B. D.

    2016-04-01

    This paper presents the results of the conceptual design of passive safety systems for reactor power 225 MWth using Pb-Bi coolant. Main purpose of this research is to design of heat removal system from the reactor wall. The heat from the reactor wall is removed by RVACS system using the natural circulation from the atmosphere around the reactor at steady state. The calculation is performed numerically using Newton-Raphson method. The analysis involves the heat transfer systems in a radiation, conduction and natural convection. Heat transfer calculations is performed on the elements of the reactor vessel, outer wall of guard vessel and the separator plate. The simulation results conclude that the conceptual design is able to remove heat 1.33% to 4.67% from the thermal reactor power. It’s can be hypothesized if the reactor had an accident, the system can still overcome the heat due to decay.

  11. Optimal Coupling of a Nuclear Reactor and a Thermal Desalination Plant

    SciTech Connect

    Caruso, G.; Naviglio, A.; Nisan, S.; Bielak, B.; Cinotti, L.; Humphries, J.R.; Martins, N.; Volpi, L.

    2002-07-01

    The present study, performed in the framework of the EURODESAL Project (5. EU FWP), deals with the analysis of the 'optimum' coupling of a PWR and of a HTGR plant with a thermal desalination plant, based on the Multiple Effects process. The reference reactors are the AP600 and the PWR900 as Pressurized reactors and the GT-MHR as Gas reactor. The calculations performed show that there are several technical solutions allowing to couple PWRs and GRs to a ME desalination plant. The optimization criteria concern the technical feasibility of the coupling, producing the maximum quantity of fresh water at the lower cost, without unacceptable reduction of the electrical power produced and without undue health hazard for population. (authors)

  12. 78 FR 59981 - Proposed Revision to Physical Security-Standard Design Certification and Operating Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-30

    ... COMMISSION Proposed Revision to Physical Security--Standard Design Certification and Operating Reactors...: Section 13.6.2 ``Physical Security--Standard Design Certification and Operating Reactors.'' The NRC seeks... security reviews of design certification applications. DATES: Comments must be filed no later than...

  13. Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.

    2007-01-01

    In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is

  14. Goals of an Undergraduate Plant Design Course.

    ERIC Educational Resources Information Center

    Baasel, William D.

    1982-01-01

    Students complete a project at Ohio University to understand the process of plant design. This and other goals of a plant design course are discussed, including student/instructor presentations and typical problems confronted by the instructors of the course. (JN)

  15. Design Of Feedforward Controllers For Multivariable Plants

    NASA Technical Reports Server (NTRS)

    Seraji, Homayoun

    1989-01-01

    Controllers based on simple low-order transfer functions. Mathematical criteria derived for design of feedforward controllers for class of multiple-input/multiple-output linear plants. Represented by simple low-order transfer functions, obtained without reconstruction of states of commands and disturbances. Enables plant to track command while remaining unresponsive to disturbance in steady state. Feedback controller added independently to stabilize plant or to make control system less susceptible to variations in parameters of plant.

  16. DOE small scale fuel alcohol plant design

    SciTech Connect

    LaRue, D.M.; Richardson, J.G.

    1980-01-01

    The Department of Energy, in an effort to facilitate the deployment of rural-based ethanol production capability, has undertaken this effort to develop a basic small-scale plant design capable of producing anhydrous ethanol. The design, when completed, will contain all necessary specifications and diagrams sufficient for the construction of a plant. The design concept is modular; that is, sections of the plant can stand alone or be integrated into other designs with comparable throughput rates. The plant design will be easily scaled up or down from the designed flow rate of 25 gallons of ethanol per hour. Conversion factors will be provided with the final design package to explain scale-up and scale-down procedures. The intent of this program is to provide potential small-scale producers with sound information about the size, engineering requirements, costs and level of effort in building such a system.

  17. Predictive Modeling in Plasma Reactor and Process Design

    NASA Technical Reports Server (NTRS)

    Hash, D. B.; Bose, D.; Govindan, T. R.; Meyyappan, M.; Arnold, James O. (Technical Monitor)

    1997-01-01

    Research continues toward the improvement and increased understanding of high-density plasma tools. Such reactor systems are lauded for their independent control of ion flux and energy enabling high etch rates with low ion damage and for their improved ion velocity anisotropy resulting from thin collisionless sheaths and low neutral pressures. Still, with the transition to 300 mm processing, achieving etch uniformity and high etch rates concurrently may be a formidable task for such large diameter wafers for which computational modeling can play an important role in successful reactor and process design. The inductively coupled plasma (ICP) reactor is the focus of the present investigation. The present work attempts to understand the fundamental physical phenomena of such systems through computational modeling. Simulations will be presented using both computational fluid dynamics (CFD) techniques and the direct simulation Monte Carlo (DSMC) method for argon and chlorine discharges. ICP reactors generally operate at pressures on the order of 1 to 10 mTorr. At such low pressures, rarefaction can be significant to the degree that the constitutive relations used in typical CFD techniques become invalid and a particle simulation must be employed. This work will assess the extent to which CFD can be applied and evaluate the degree to which accuracy is lost in prediction of the phenomenon of interest; i.e., etch rate. If the CFD approach is found reasonably accurate and bench-marked with DSMC and experimental results, it has the potential to serve as a design tool due to the rapid time relative to DSMC. The continuum CFD simulation solves the governing equations for plasma flow using a finite difference technique with an implicit Gauss-Seidel Line Relaxation method for time marching toward a converged solution. The equation set consists of mass conservation for each species, separate energy equations for the electrons and heavy species, and momentum equations for the gas

  18. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  19. A tritium permeation model for conceptual fusion reactor designs

    NASA Astrophysics Data System (ADS)

    Hanchar, D. R.; Kazimi, M. S.

    1983-02-01

    A transient tritium permeation model is developed based on a simplified conceptual DT-fueled fusion reactor design. The major design features described in the model are a solid breeder blanket, a low pressure purge gas in the blanket, and a high pressure helium primary coolant. Tritium inventory in the breeder is considered to be due to diffusive hold-up and solubility effects. It is assumed that diffusive hold-up is the dominant factor in order to separate the solution for the breeder tritium concentration. The model was applied to the STARFIRE-Interim Reference Design, whose system parameters yielded a breeder tritium inventory on the order of grams, based on an average pellet radius of 10-3 cm. The breeder pellets reach their steady-state tritium content in approximately 1.4×104 s from system start-up, assuming continuous full power operation. Both the steady-state breeder tritium concentration and the time to reach that steady-state are proportional to the pellet radius squared. Other candidate solid breeders were considered, and their effect on the blanket tritium inventory was noted. The addition of oxygen to the primary coolant loop was required in order to keep the tritium losses through the heat exchanger to within the design goal of 0.1 Ci/day.

  20. Design of GA thermochemical water-splitting process for the Mirror Advanced Reactor System

    SciTech Connect

    Brown, L.C.

    1983-04-01

    GA interfaced the sulfur-iodine thermochemical water-splitting cycle to the Mirror Advanced Reactor System (MARS). The results of this effort follow as one section and part of a second section to be included in the MARS final report. This section describes the process and its interface to the reactor. The capital and operating costs for the hydrogen plant are described.

  1. Design Status and Applications of Small Reactors Without On-Site Refuelling

    SciTech Connect

    Kuznetsov, Vladimir

    2006-07-01

    Small reactors without on-site refuelling (SRWORs) are the reactors that can operate without reloading and shuffling of fuel for a reasonably long period with no refuelling equipment being present in the reactor and no fuel being stored at the site during reactor operation. By virtue of being small, transportable and requiring no operations with fuel from a customer, such reactors form an attractive domain for fuel or even NPP leasing. SRWORs could simplify the implementation of safeguards and provide certain guarantees of sovereignty to those countries that would agree to forego the development of the indigenous fuel cycle. About 30 concepts of such reactors are being analyzed or developed in 6 IAEA member states. Based on intermediate results of IAEA activities in support of the design and technology development for such reactors, the paper provides technical details on the design status, fuel cycle options and possible applications of SRWORs. (authors)

  2. Standardizing electrocoagulation reactor design: iron electrodes for NOM removal.

    PubMed

    Dubrawski, Kristian L; Mohseni, Madjid

    2013-03-01

    A novel systematic approach for reactor design was described for iron electrocoagulation (EC) and applied to drinking water treatment. Suwannee NOM was used as a model compound; performance was quantified by UV-abs-254 and DOC removal. Significant EC design variables were identified and examined: current density (i) (2.43-26.8 mA cm(-2)), coagulant or charge loading rate (CLR) (100-1000 CL(-1) min(-1)), and flocculation methodology ("fast" and "slow"). A correlation was found between increased i and decreased current efficiency (φ), optimum NOM removal was found at i ~10 mA cm(-2). A lower CLR showed greater total DOC removal, while a higher CLR led to less reactor residence time and required either longer flocculation times or greater coagulant dose for similar NOM removal. This paper defines and describes the four general EC "classes" of operation that have implications on several important measures of success: coagulant dose, electrical consumption, process speed, volumetric footprint, and post-EC flocculation requirements. Two classes were further examined with or without pH adjustment for DOC removal, showing that a "fast" EC mode without flocculation is more appropriate for smaller applications, while a "slow" EC mode is more effective for large permanent applications, where flocculation and settling can reduce coagulant and electrical consumption. The effect of pH adjustment showed greater impact with the "fast" dosing mode than with the "slow" mode, adjustment to pH 6 with the "fast" mode gave 13.8% and 29.1% greater DOC and UV-abs-254 removal, respectively, compared to the baseline without pH adjustment.

  3. Boiling water neutronic reactor incorporating a process inherent safety design

    DOEpatents

    Forsberg, Charles W.

    1987-01-01

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  4. Boiling water neutronic reactor incorporating a process inherent safety design

    DOEpatents

    Forsberg, C.W.

    1985-02-19

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  5. OSIRIS and SOMBRERO Inertial Fusion Power Plant Designs, Volume 1: Executive Summary & Overview

    SciTech Connect

    Meier, W. R.; Bieri, R. L.; Monsler, M. J.; Hendricks, C.D.; Laybourne, P.; Shillito, K. R.

    1992-03-01

    This is a comprehensive design study of two Inertial Fusion Energy (IFE) electric power plants. Conceptual designs are presented for a fusion reactor (called Osiris) using an induction-linac heavy-ion beam driver, and another (called SOMBRERO) using a KrF laser driver. The designs covered all aspects of IFE power plants, including the chambers, heat transport and power conversion systems, balance-of-plant facilities, target fabrication, target injection and tracking, as well as the heavy-ion and KrF drivers. The point designs were assessed and compared in terms of their environmental & safety aspects, reliability and availability economics, and technology development needs.

  6. 10 CFR Appendix N to Part 52 - Standardization of Nuclear Power Plant Designs: Combined Licenses To Construct and Operate...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Standardization of Nuclear Power Plant Designs: Combined Licenses To Construct and Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... FOR NUCLEAR POWER PLANTS Pt. 52, App. N Appendix N to Part 52—Standardization of Nuclear Power...

  7. 10 CFR Appendix N to Part 52 - Standardization of Nuclear Power Plant Designs: Combined Licenses To Construct and Operate...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Standardization of Nuclear Power Plant Designs: Combined Licenses To Construct and Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... FOR NUCLEAR POWER PLANTS Pt. 52, App. N Appendix N to Part 52—Standardization of Nuclear Power...

  8. 10 CFR Appendix N to Part 52 - Standardization of Nuclear Power Plant Designs: Combined Licenses To Construct and Operate...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Standardization of Nuclear Power Plant Designs: Combined Licenses To Construct and Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... FOR NUCLEAR POWER PLANTS Pt. 52, App. N Appendix N to Part 52—Standardization of Nuclear Power...

  9. 10 CFR Appendix N to Part 52 - Standardization of Nuclear Power Plant Designs: Combined Licenses To Construct and Operate...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Standardization of Nuclear Power Plant Designs: Combined Licenses To Construct and Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... FOR NUCLEAR POWER PLANTS Pt. 52, App. N Appendix N to Part 52—Standardization of Nuclear Power...

  10. Heat Pipe Space Nuclear Reactor Design Assessment. Volume 1. Design Status of the SP-100 Heat Pipe Space Nuclear Reactor System

    DTIC Science & Technology

    1985-08-01

    design studies for a higher power , unmanned nuclear reactor space power source with a long design lifetime (7 yr). A 100 kWe , high temperature...reactors are potentially the best source of high power levels (in excess of 100 kWe ) for space. Other possible sources are chemical combustion...configured as a com- pact power source of greater than 10 kWe . Table 1 compares solar to nuclear power at three

  11. Separation Requirements for a Hydrogen Production Plant and High-Temperature Nuclear Reactor

    SciTech Connect

    Curtis Smith; Scott Beck; Bill Galyean

    2005-09-01

    This report provides the methods, models, and results of an evaluation for locating a hydrogen production facility near a nuclear power plant. In order to answer the risk-related questions for this combined nuclear and chemical facility, we utilized standard probabilistic safety assessment methodologies to answer three questions: what can happen, how likely is it, and what are the consequences? As part of answering these questions, we developed a model suitable to determine separation distances for hydrogen process structures and the nuclear plant structures. Our objective of the model-development and analysis is to answer key safety questions related to the placement of one or more hydrogen production plants in the vicinity of a high-temperature nuclear reactor. From a thermal-hydraulic standpoint we would like the two facilities to be quite close. However, safety and regulatory implications force the separation distance to be increased, perhaps substantially. Without answering these safety questions, the likelihood for obtaining a permit to construct and build such as facility in the U.S. would be questionable. The quantitative analysis performed for this report provides us with a scoping mechanism to determine key parameters related to the development of a nuclear-based hydrogen production facility. From our calculations, we estimate that when the separation distance is less than 100m, the core damage frequency is large enough (greater than 1E-6/yr) to become problematic in a risk-informed environment. However, a variety of design modifications, for example blast-deflection barriers, were explored to determine the impact of potential mitigating strategies. We found that these mitigating cases may significantly reduce risk and should be explored as the design for the hydrogen production facility evolves.

  12. High Flux Isotope Reactor cold neutron source reference design concept

    SciTech Connect

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

    1998-05-01

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

  13. TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report

    SciTech Connect

    Conn, R.W.; Ghoniem, N.M.; Firestone, M.A.

    1986-09-01

    Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m/sup 2/ and a surface heat flux of 1 MW/m/sup 2/. The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO/sub 2/ rods. The helium coolant pressure is 5 MPa, entering the module at 297/sup 0/C and exiting at 550/sup 0/C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter.

  14. A brief history of design studies on innovative nuclear reactors

    SciTech Connect

    Sekimoto, Hiroshi

    2014-09-30

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  15. ABWR (advanced boiling water reactor) Design Verification Program

    SciTech Connect

    Fox, J.N.

    1990-10-01

    The ABWR Design Verification Program is aimed at restoring confidence in the US licensing process by demonstrating its workability by obtaining USNRC preapproval of GE's ABWR Standard Plant. The purpose of this work is to achieve full NRC approval of the ABWR through the award of an NRC Staff final design approval (FDA) and design certification. The approach is to (1) establish a licensing basis with the NRC Staff for the ABWR, (2) prepare and submit, for NRC Staff review, an SSAR to obtain an FDA, and (3) participate in a rulemaking process to obtain certification of the ABWR design. This program was initiated August 27, 1986. This report, the fourth annual progress report, summarizes progress on this program from October 1, 1989 through September 30, 1990. 9 refs., 5 tabs.

  16. Comparison of oxide- and metal-core behavior during CRBRP (Clinch River Breeder Reactor Plant) station blackout

    SciTech Connect

    Polkinghorne, S T; Atkinson, S A

    1986-01-01

    A resurrected concept that could significantly improve the inherently safe response of Liquid-Metal cooled Reactors (LMRs) during severe undercooling transients is the use of metallic fuel. Analytical studies have been reported on for the transient behavior of metal-fuel cores in innovative, inherently safe LMR designs. This paper reports on an analysis done, instead, for the Clinch River Breeder Reactor Plant (CRBRP) design with the only innovative change being the incorporation of a metal-fuel core. The SSC-L code was used to simulate a protected station blackout accident in the CRBRP with a 943 MWt Integral Fast Reactor (IFR) metal-fuel core. The results, compared with those for the oxide-fueled CRBRP, show that the margin to boiling is greater for the IFR core. However, the cooldown transient is more severe due to the faster thermal response time of metallic fuel. Some additional calculations to assess possible LMR design improvements (reduced primary system pressure losses, extended flow coastdown) are also discussed. 8 refs., 13 figs., 2 tabs.

  17. NGNP: High Temperature Gas-Cooled Reactor Key Definitions, Plant Capabilities, and Assumptions

    SciTech Connect

    Wayne Moe

    2013-05-01

    This document provides key definitions, plant capabilities, and inputs and assumptions related to the Next Generation Nuclear Plant to be used in ongoing efforts related to the licensing and deployment of a high temperature gas-cooled reactor. These definitions, capabilities, and assumptions were extracted from a number of NGNP Project sources such as licensing related white papers, previously issued requirement documents, and preapplication interactions with the Nuclear Regulatory Commission (NRC).

  18. Helium turbomachine design for GT-MHR power plant

    SciTech Connect

    McDonald, C.F.; Orlando, R.J.; Cotzas, G.M.

    1994-07-01

    The power conversion system in the gas turbine modular helium reactor (GT-MHR) power plant is based on a highly recuperated closed Brayton cycle. The major component in the direct cycle system is a helium closed-cycle gas turbine rated at 286 MW(e). The rotating group consists of an intercooled helium turbocompressor coupled to a synchronous generator. The vertical rotating assembly is installed in a steel vessel, together with the other major components (i.e., recuperator, precooler, intercooler, and connecting ducts and support structures). The rotor is supported on an active magnetic bearing system. The turbine operates directly on the reactor helium coolant, and with a temperature of 850{degree}C (1562{degree}F) the plant efficiency is over 47%. This paper addresses the design and development planning of the helium turbomachine, and emphasizes that with the utilization of proven technology, this second generation nuclear power plant could be in service in the first decade of the 21st century.

  19. Design Option of Heat Exchanger for the Next Generation Nuclear Plant

    SciTech Connect

    Eung Soo Kim; Chang Oh

    2008-09-01

    The Next Generation Nuclear Plant (NGNP), a very High temperature Gas-Cooled Reactor (VHTGRS) concept, will provide the first demonstration of a closed-loop Brayton cycle at a commercial scale of a few hundred megawatts electric and hydrogen production. The power conversion system (PCS) for the NGNP will take advantage of the significantly higher reactor outlet temperatures of the VHTGRS to provide higher efficiencies than can be achieved in the current generation of light water reactors. Besides demonstrating a system design that can be used directly for subsequent commercial deployment, the NGNP will demonstrate key technology elements that can be used in subsequent advanced power conversion systems for other Generation IV reactors. In anticipation of the design, development and procurement of an advanced power conversion system for the NGNP, the system integration of the NGNP and hydrogen plant was initiated to identify the important design and technology options that must be considered in evaluating the performance of the proposed NGNP. As part of the system integration of the VHTGRS and hydrogen production plant, the intermediate heat exchanger is used to transfer the process heat from VHTGRS to hydrogen plant. Therefore, the design and configuration of the intermediate heat exchanger are very important. This paper will include analysis of one stage versus two stage heat exchanger design configurations and thermal stress analyses of a printed circuit heat exchanger, helical coil heat exchanger, and shell/tube heat exchanger.

  20. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  1. Conceptual physics design of a compact torus fusion reactor (CTOR)

    SciTech Connect

    Hagenson, R.L.; Krakowski, R.A.

    1980-07-01

    The general approach to fusion power embodied by field-reversed plasmoid configurations is reviewed within the context of a power reactor. A simple analytic formulation is developed and applied to the field-reversed theta pinch as a core plasma for a thermonuclear reactor. These calculations and results are based on a minimum power constraint and will serve as a basis for more exact and detailed reactor modeling.

  2. Virtual environments for nuclear power plant design

    SciTech Connect

    Brown-VanHoozer, S.A.; Singleterry, R.C. Jr.; King, R.W.

    1996-03-01

    In the design and operation of nuclear power plants, the visualization process inherent in virtual environments (VE) allows for abstract design concepts to be made concrete and simulated without using a physical mock-up. This helps reduce the time and effort required to design and understand the system, thus providing the design team with a less complicated arrangement. Also, the outcome of human interactions with the components and system can be minimized through various testing of scenarios in real-time without the threat of injury to the user or damage to the equipment. If implemented, this will lead to a minimal total design and construction effort for nuclear power plants (NPP).

  3. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    SciTech Connect

    1995-08-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

  4. An Innovative Hybrid Loop-Pool Design for Sodium Cooled Fast Reactor

    SciTech Connect

    Haihua Zhao; Hongbin Zhang

    2007-11-01

    The existing sodium cooled fast reactors (SFR) have two types of designs – loop type and pool type. In the loop type design, such as JOYO (Japan) [1] and MONJU (Japan), the primary coolant is circulated through intermediate heat exchangers (IHX) external to the reactor tank. The major advantages of loop design include compactness and easy maintenance. The disadvantage is higher possibility of sodium leakage. In the pool type design such as EBR-II (USA), BN-600M(Russia), Superphénix (France) and European Fast Reactor [2], the reactor core, primary pumps, IHXs and direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) all are immersed in a pool of sodium coolant within the reactor vessel, making a loss of primary coolant extremely unlikely. However, the pool type design makes primary system large. In the latest ANL’s Advanced Burner Test Reactor (ABTR) design [3], the primary system is configured in a pool-type arrangement. The hot sodium at core outlet temperature in hot pool is separated from the cold sodium at core inlet temperature in cold pool by a single integrated structure called Redan. Redan provides the exchange of the hot sodium from hot pool to cold pool through IHXs. The IHXs were chosen as the traditional tube-shell design. This type of IHXs is large in size and hence large reactor vessel is needed.

  5. High flux isotope reactor cold source preconceptual design study report

    SciTech Connect

    Selby, D.L.; Bucholz, J.A.; Burnette, S.E.

    1995-12-01

    In February 1995, the deputy director of Oak Ridge National Laboratory (ORNL) formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced Neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. The anticipated cold source will consist of a cryogenic LH{sub 2} moderator plug, a cryogenic pump system, a refrigerator that uses helium gas as a refrigerant, a heat exchanger to interface the refrigerant with the hydrogen loop, liquid hydrogen transfer lines, a gas handling system that includes vacuum lines, and an instrumentation and control system to provide constant system status monitoring and to maintain system stability. The scope of this project includes the development, design, safety analysis, procurement/fabrication, testing, and installation of all of the components necessary to produce a working cold source within an existing HFIR beam tube. This project will also include those activities necessary to transport the cold neutron beam to the front face of the present HFIR beam room. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and research and development (R and D), (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the preconceptual phase and establishes the concept feasibility. The information presented includes the project scope, the preliminary design requirements, the preliminary cost and schedule, the preliminary performance data, and an outline of the various plans for completing the project.

  6. Plant Growth Module (PGM) conceptual design

    NASA Technical Reports Server (NTRS)

    Schwartzkopf, Steven H.; Rasmussen, Daryl

    1987-01-01

    The Plant Growth Module for the Controlled Ecological Life Support System (CELSS), designed to answer basic science questions related to growing plants in closed systems, is described functionally with artist's conception drawings. Subsystems are also described, including enclosure and access; data acquisition and control; gas monitor and control; heating, ventilation, and air conditioning; air delivery; nutrient monitor and control; microbial monitoring and control; plant support and nutrient delivery; illumination; and internal operations. The hardware development plan is outlined.

  7. Reactor Physics Parametric and Depletion Studies in Support of TRISO Particle Fuel Specification for the Next Generation Nuclear Plant

    SciTech Connect

    James W. Sterbentz; Bren Phillips; Robert L. Sant; Gray S. Chang; Paul D. Bayless

    2003-09-01

    Reactor physics calculations were initiated to answer several major questions related to the proposed TRISO-coated particle fuel that is to be used in the prismatic Very High Temperature Reactor (VHTR) or the Next Generation Nuclear Plant (NGNP). These preliminary design evaluation calculations help ensure that the upcoming fuel irradiation tests will test appropriate size and type of fuel particles for a future NGNP reactor design. Conclusions from these calculations are expected to confirm and suggest possible modifications to the current particle fuel parameters specified in the evolving Fuel Specification. Calculated results dispel the need for a binary fuel particle system, which is proposed in the General Atomics GT-MHR concept. The GT-MHR binary system is composed of both a fissile and fertile particle with 350- and 500- micron kernel diameters, respectively. For the NGNP reactor, a single fissile particle system (single UCO kernel size) can meet the reactivity and power cycle length requirements demanded of the NGNP. At the same time, it will provide substantial programmatic cost savings by eliminating the need for dual particle fabrication process lines and dual fuel particle irradiation tests required of a binary system. Use of a larger 425-micron kernel diameter single fissile particle (proposed here), as opposed to the 350-micron GT-MHR fissile particle size, helps alleviate current compact particle packing fractions fabrication limitations (<35%), improves fuel block loading for higher n-batch reload options, and tracks the historical correlation between particle size and enrichment (10 and 14 wt% U-235 particle enrichments are proposed for the NGNP). Overall, the use of the slightly larger kernel significantly broadens the NGNP reactor core design envelope and provides increased design margin to accommodate the (as yet) unknown final NGNP reactor design. Maximum power-peaking factors are calculated for both the initial and equilibrium NGNP cores

  8. Comparison of reactor designs for biological ex-situ soil treatment

    SciTech Connect

    Potter, C.; Haught, R.

    1995-10-01

    The goal of this project is to evaluate the potential of biological ex-situ soil treatment systems to remediate soils contaminated with hazardous chemicals. A laminar-type flow pilot-scale reactor with volume of 3 cu. yd has been constructed at EPA`s Test & Evaluation J&D Facility in Cincinnati. Laminar-type flow from one side of the reactor to the other may provide even aeration to all areas of the reactor while avoiding the use of pipes inside the reactor. This design greatly facilitates loading and unloading the reactor and is readily scalable to larger systems. Passing smoke through the reactor for visual observation of flow indicated uniform flow in the empty reactor. Further testing involves filling the reactor with vermiculite, flushing with Argon, and then passing air through the reactor at about 1 volume change per day to evaluate air flow through this uniform solid matrix. Oxygen probes, located at 27 positions within the reactor, provide information on air progression through the system. Analysis of gas flow through an empty reactor and through a uniform matrix (vermiculite) allows separation of reactor flaws from soil inhomogeneities as causes of any nonuniform aeration of the reactor space. Soil contaminated with polynuclear aromatic hydrocarbons (PAHs) from The Reilly Tar Pit Superfund Site in St. Louis Park, MN has been shipped under the small quantity treatability exemption to the T&E Facility for research on soil aeration and effectiveness of this ex-situ reactor design for biological treatment of contaminated soils.

  9. A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

    SciTech Connect

    John Darrell Bess

    2008-06-01

    A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for

  10. Investigation of plant control strategies for the supercritical C0{sub 2}Brayton cycle for a sodium-cooled fast reactor using the plant dynamics code.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J.

    2011-04-12

    The development of a control strategy for the supercritical CO{sub 2} (S-CO{sub 2}) Brayton cycle has been extended to the investigation of alternate control strategies for a Sodium-Cooled Fast Reactor (SFR) nuclear power plant incorporating a S-CO{sub 2} Brayton cycle power converter. The SFR assumed is the 400 MWe (1000 MWt) ABR-1000 preconceptual design incorporating metallic fuel. Three alternative idealized schemes for controlling the reactor side of the plant in combination with the existing automatic control strategy for the S-CO{sub 2} Brayton cycle are explored using the ANL Plant Dynamics Code together with the SAS4A/SASSYS-1 Liquid Metal Reactor (LMR) Analysis Code System coupled together using the iterative coupling formulation previously developed and implemented into the Plant Dynamics Code. The first option assumes that the reactor side can be ideally controlled through movement of control rods and changing the speeds of both the primary and intermediate coolant system sodium pumps such that the intermediate sodium flow rate and inlet temperature to the sodium-to-CO{sub 2} heat exchanger (RHX) remain unvarying while the intermediate sodium outlet temperature changes as the load demand from the electric grid changes and the S-CO{sub 2} cycle conditions adjust according to the S-CO{sub 2} cycle control strategy. For this option, the reactor plant follows an assumed change in load demand from 100 to 0 % nominal at 5 % reduction per minute in a suitable fashion. The second option allows the reactor core power and primary and intermediate coolant system sodium pump flow rates to change autonomously in response to the strong reactivity feedbacks of the metallic fueled core and assumed constant pump torques representing unchanging output from the pump electric motors. The plant behavior to the assumed load demand reduction is surprising close to that calculated for the first option. The only negative result observed is a slight increase in the intermediate

  11. Considerations Associated with Reactor Technology Selection for the Next Generation Nuclear Plant Project

    SciTech Connect

    L.E. Demick

    2010-09-01

    At the inception of the Next Generation Nuclear Plant Project and during predecessor activities, alternative reactor technologies have been evaluated to determine the technology that best fulfills the functional and performance requirements of the targeted energy applications and market. Unlike the case of electric power generation where the reactor performance is primarily expressed in terms of economics, the targeted energy applications involve industrial applications that have specific needs in terms of acceptable heat transport fluids and the associated thermodynamic conditions. Hence, to be of interest to these industrial energy applications, the alternative reactor technologies are weighed in terms of the reactor coolant/heat transport fluid, achievable reactor outlet temperature, and practicality of operations to achieve the very high reliability demands associated with the petrochemical, petroleum, metals and related industries. These evaluations have concluded that the high temperature gas-cooled reactor (HTGR) can uniquely provide the required ranges of energy needs for these target applications, do so with promising economics, and can be commercialized with reasonable development risk in the time frames of current industry interest – i.e., within the next 10-15 years.

  12. What went Right: Resilience of Existing Reactors to - for Generation III+ Reactor Design

    NASA Astrophysics Data System (ADS)

    Garwin, Richard L.

    2014-07-01

    To quote Tolstoy's Anna Karenina, "All happy families are alike; each unhappy family is unhappy in its own way." So the reactors that have been working well in the world don't get a lot of attention...

  13. Materials compatibility considerations for a fusion-fission hybrid reactor design

    SciTech Connect

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

    1983-01-01

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

  14. Coagulant recovery from water treatment plant sludge and reuse in post-treatment of UASB reactor effluent treating municipal wastewater.

    PubMed

    Nair, Abhilash T; Ahammed, M Mansoor

    2014-09-01

    In the present study, feasibility of recovering the coagulant from water treatment plant sludge with sulphuric acid and reusing it in post-treatment of upflow anaerobic sludge blanket (UASB) reactor effluent treating municipal wastewater were studied. The optimum conditions for coagulant recovery from water treatment plant sludge were investigated using response surface methodology (RSM). Sludge obtained from plants that use polyaluminium chloride (PACl) and alum coagulant was utilised for the study. Effect of three variables, pH, solid content and mixing time was studied using a Box-Behnken statistical experimental design. RSM model was developed based on the experimental aluminium recovery, and the response plots were developed. Results of the study showed significant effects of all the three variables and their interactions in the recovery process. The optimum aluminium recovery of 73.26 and 62.73 % from PACl sludge and alum sludge, respectively, was obtained at pH of 2.0, solid content of 0.5 % and mixing time of 30 min. The recovered coagulant solution had elevated concentrations of certain metals and chemical oxygen demand (COD) which raised concern about its reuse potential in water treatment. Hence, the coagulant recovered from PACl sludge was reused as coagulant for post-treatment of UASB reactor effluent treating municipal wastewater. The recovered coagulant gave 71 % COD, 80 % turbidity, 89 % phosphate, 77 % suspended solids and 99.5 % total coliform removal at 25 mg Al/L. Fresh PACl also gave similar performance but at higher dose of 40 mg Al/L. The results suggest that coagulant can be recovered from water treatment plant sludge and can be used to treat UASB reactor effluent treating municipal wastewater which can reduce the consumption of fresh coagulant in wastewater treatment.

  15. Osiris and SOMBRERO inertial confinement fusion power plant designs. Volume 2, Designs, assessments, and comparisons, Final report

    SciTech Connect

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

    1992-03-01

    The primary objective of the of the IFE Reactor Design Studies was to provide the Office of Fusion Energy with an evaluation of the potential of inertial fusion for electric power production. The term reactor studies is somewhat of a misnomer since these studies included the conceptual design and analysis of all aspects of the IFE power plants: the chambers, heat transport and power conversion systems, other balance of plant facilities, target systems (including the target production, injection, and tracking systems), and the two drivers. The scope of the IFE Reactor Design Studies was quite ambitious. The majority of our effort was spent on the conceptual design of two IFE electric power plants, one using an induction linac heavy ion beam (HIB) driver and the other using a Krypton Fluoride (KrF) laser driver. After the two point designs were developed, they were assessed in terms of their (1) environmental and safety aspects; (2) reliability, availability, and maintainability; (3) technical issues and technology development requirements; and (4) economics. Finally, we compared the design features and the results of the assessments for the two designs.

  16. Design basis for protection of light water nuclear power plants against effects of postulated pipe rupture

    SciTech Connect

    Not Available

    1981-01-01

    This standard addresses the design bases for light water reactor, nuclear power plant structures and components essential for the protection of public health and safety from the potential adverse effects of pipe whip, jet impingement, pressurization of compartments outside containment, environmental conditions and flooding associated with a postulated pipe rupture. The design bases for missile protection and the design bases for containment pressurization are not within this standard.

  17. The development of the fast-running simulation pressurized water reactor plant analyzer code (NUPAC-1)

    SciTech Connect

    Sasaki, K.; Terashita, N.; Ogino, T. . Central Research Lab.)

    1989-06-01

    This article discusses a pressurized water reactor plant analyzer code (NUPAC-1) has been developed to apply to an operator support system or an advanced training simulator. The simulation code must produce reasonably accurate results as well as fun in a fast mode for realizing functions such as anomaly detection, estimation of unobservable plant internal states, and prediction of plant state trends. The NUPAC-1 code adopts fast computing methods, i.e., the table fitting method of the state variables, time-step control, and calculation control of heat transfer coefficients, in order to attain accuracy and fast-running capability.

  18. Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

    NASA Astrophysics Data System (ADS)

    Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

  19. A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment

    SciTech Connect

    Wheeler, Timothy A.; Liao, Huafei

    2014-12-01

    United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constituted a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.

  20. GPU Based General-Purpose Parallel computing to Solve Nuclear Reactor In-Core fuel Management Design and Operation Problem

    NASA Astrophysics Data System (ADS)

    Prayudhatama, D.; Waris, A.; Kurniasih, N.; Kurniadi, R.

    2010-06-01

    In-core fuel management study is a crucial activity in nuclear power plant design and operation. Its common problem is to find an optimum arrangement of fuel assemblies inside the reactor core. Main objective for this activity is to reduce the cost of generating electricity, which can be done by altering several physical properties of the nuclear reactor without violating any of the constraints imposed by operational and safety considerations. This research try to address the problem of nuclear fuel arrangement problem, which is, leads to the multi-objective optimization problem. However, the calculation of the reactor core physical properties itself is a heavy computation, which became obstacle in solving the optimization problem by using genetic algorithm optimization. This research tends to address that problem by using the emerging General Purpose Computation on Graphics Processing Units (GPGPU) techniques implemented by C language for CUDA (Compute Unified Device Architecture) parallel programming. By using this parallel programming technique, we develop parallelized nuclear reactor fitness calculation, which is involving numerical finite difference computation. This paper describes current prototype of the parallel algorithm code we have developed on CUDA, that performs one hundreds finite difference calculation for nuclear reactor fitness evaluation in parallel by using GPU G9 Hardware Series developed by NVIDIA.

  1. GPU Based General-Purpose Parallel computing to Solve Nuclear Reactor In-Core fuel Management Design and Operation Problem

    SciTech Connect

    Prayudhatama, D.; Waris, A.; Kurniasih, N.; Kurniadi, R.

    2010-06-22

    In-core fuel management study is a crucial activity in nuclear power plant design and operation. Its common problem is to find an optimum arrangement of fuel assemblies inside the reactor core. Main objective for this activity is to reduce the cost of generating electricity, which can be done by altering several physical properties of the nuclear reactor without violating any of the constraints imposed by operational and safety considerations. This research try to address the problem of nuclear fuel arrangement problem, which is, leads to the multi-objective optimization problem. However, the calculation of the reactor core physical properties itself is a heavy computation, which became obstacle in solving the optimization problem by using genetic algorithm optimization.This research tends to address that problem by using the emerging General Purpose Computation on Graphics Processing Units (GPGPU) techniques implemented by C language for CUDA (Compute Unified Device Architecture) parallel programming. By using this parallel programming technique, we develop parallelized nuclear reactor fitness calculation, which is involving numerical finite difference computation. This paper describes current prototype of the parallel algorithm code we have developed on CUDA, that performs one hundreds finite difference calculation for nuclear reactor fitness evaluation in parallel by using GPU G9 Hardware Series developed by NVIDIA.

  2. Design-development and operation of the Experimental Boiling-Water Reactor (EBWR) facility, 1955--1967

    SciTech Connect

    Boing, L.E.; Wimunc, E.A.; Whittington, G.A.

    1990-11-01

    The Experimental Boiling-Water Reactor (EBWR) was designed, built, and operated to provide experience and engineering data that would demonstrate the feasibility of the direct-cycle, boiling-water reactor and be applicable to improved, larger nuclear power stations; and was based on information obtained in the first test boiling-water reactors, the BORAX series. EBWR initially produced 20 MW(t), 5 MW(e); later modified and upgraded, as described and illustrated, it was operated at up to 100 MW(t). The facility fulfilled its primary mission -- demonstrating the practicality of the direct-boiling concept -- and, in fact, was the prototype of some of the first commercial plants and of reactor programs in some other countries. After successful completion of the Water-Cooled Reactor Program, EBWR was utilized in the joint Argonne-Hanford Plutonium Recycle Program to develop data for the utilization of plutonium as a fuel in light- water thermal systems. Final shutdown of the EBWR facility followed the termination of the latter program. 13 refs., 12 figs.

  3. Wastewater treatment with submerged fixed bed biofilm reactor systems--design rules, operating experiences and ongoing developments.

    PubMed

    Schlegel, S; Koeser, H

    2007-01-01

    Wastewater treatment systems using bio-films that grow attached to a support media are an alternative to the widely used suspended growth activated sludge process. Different fixed growth biofilm reactors are commercially used for the treatment of municipal as well as industrial wastewater. In this paper a fairly new fixed growth biofilm system, the submerged fixed bed biofilm reactor (SFBBR), is discussed. SFBBRs are based on aerated submerged fixed open structured plastic media for the support of the biofilm. They are generally operated without sludge recirculation in order to avoid clogging of the support media and problems with the control of the biofilm. Reactor and process design considerations for these reactors are reviewed. Measures to ensure the development and maintenance of an active biofilm are examined. SFBBRs have been applied successfully to small wastewater treatment plants where complete nitrification but no high degree of denitrification is necessary. For the pre-treatment of industrial wastewater the use of SFBBRs is advantageous, especially in cases of wastewater with high organic loading or high content of compounds with low biodegradability. Performance data from exemplary commercial plants are given. Ongoing research and development efforts aim at achieving a high simultaneous total nitrogen (TN) removal of aerated SFBBRs and at improving the efficiency of TN removal in anoxic SFBBRs.

  4. LBB considerations for a new plant design

    SciTech Connect

    Swamy, S.A.; Mandava, P.R.; Bhowmick, D.C.; Prager, D.E.

    1997-04-01

    The leak-before-break (LBB) methodology is accepted as a technically justifiable approach for eliminating postulation of Double-Ended Guillotine Breaks (DEGB) in high energy piping systems. This is the result of extensive research, development, and rigorous evaluations by the NRC and the commercial nuclear power industry since the early 1970s. The DEGB postulation is responsible for the many hundreds of pipe whip restraints and jet shields found in commercial nuclear plants. These restraints and jet shields not only cost many millions of dollars, but also cause plant congestion leading to reduced reliability in inservice inspection and increased man-rem exposure. While use of leak-before-break technology saved hundreds of millions of dollars in backfit costs to many operating Westinghouse plants, value-impacts resulting from the application of this technology for future plants are greater on a per plant basis. These benefits will be highlighted in this paper. The LBB technology has been applied extensively to high energy piping systems in operating plants. However, there are differences between the application of LBB technology to an operating plant and to a new plant design. In this paper an approach is proposed which is suitable for application of LBB to a new plant design such as the Westinghouse AP600. The approach is based on generating Bounding Analyses Curves (BAC) for the candidate piping systems. The general methodology and criteria used for developing the BACs are based on modified GDC-4 and Standard Review Plan (SRP) 3.6.3. The BAC allows advance evaluation of the piping system from the LBB standpoint thereby assuring LBB conformance for the piping system. The piping designer can use the results of the BACs to determine acceptability of design loads and make modifications (in terms of piping layout and support configurations) as necessary at the design stage to assure LBB for the, piping systems under consideration.

  5. Dosimetry assessments for the reactor pressure vessel and core barrel in UK PWR plant

    SciTech Connect

    Thornton, D.A.; Allen, D.A.; Huggon, A.P.; Picton, D.J.; Robinson, A.T.; Steadman, R.J.; Seren, T.; Lipponen, M.; Kekki, T.

    2011-07-01

    Specimens for the Sizewell B reactor pressure vessel (RPV) inservice steels surveillance program are irradiated inside eight capsules located within the reactor pressure vessel and loaded prior to commissioning. The periodic removal of these capsules and testing of their contents provides material properties data at intervals during the lifetime of the plant. Neutron activation measurements and radiation transport calculations play an essential role in assessing the neutron exposure of the specimens and RPV. Following the most recent withdrawal, seven capsules have now been removed covering nine cycles of reactor operation. This paper summarizes the dosimetry results of the Sizewell B surveillance program obtained to date. In addition to an overview of the calculational methodology it includes a review of the measurements. Finally, it describes an extension of the methodology to provide dosimetry recommendations for the core barrel and briefly discusses the results that were obtained. (authors)

  6. Design requirements for innovative homogeneous reactor, lesson learned from Fukushima accident

    NASA Astrophysics Data System (ADS)

    Arbie, Bakri; Pinem, Suryan; Sembiring, Tagor; Subki, Iyos

    2012-06-01

    The Fukushima disaster is the largest nuclear accident since the 1986 Chernobyl disaster, but it is more complex as multiple reactors and spent fuel pools are involved. The severity of the nuclear accident is rated 7 in the International Nuclear Events Scale. Expert said that "Fukushima is the biggest industrial catastrophe in the history of mankind". According to Mitsuru Obe, in The Wall Street Journal, May 16th of 2011, TEPCO estimates the nuclear fuel was exposed to the air less than five hours after the earthquake struck. Fuel rods melted away rapidly as the temperatures inside the core reached 2800 C within six hours. In less than 16 hours, the reactor core melted and dropped to the bottom of the pressure vessel. The information should be evaluated in detail. In Germany several nuclear power plant were shutdown, Italy postponed it's nuclear power program and China reviewed their nuclear power program. Different news come from Britain, in October 11, 2011, the Safety Committee said all clear for nuclear power in Britain, because there are no risk of strong earthquake and tsunami in the region. Due to this severe fact, many nuclear scientists and engineer from all over the world are looking for a new approach, such as homogeneous reactor which was developed in Oak Ridge National Laboratory in 1960-ies, during Dr. Alvin Weinberg tenure as the Director of ORNL. The paper will describe the design requirement that will be used as the basis for innovative homogeneous reactor. Innovative Homogeneous Reactor is expected to reduce core melt by two decades (4), since the fuel is intermix homogeneously with coolant and secondly we eliminate the used fuel rod which need to be cooled for a long period of time. In order to be successful for its implementation of the innovative system, testing and validation, three phases of development will be introduced. The first phase is Low Level Goals is really the proof of concept;the Medium Level Goal is Technical Goalsand the High

  7. Production of Advanced Biofuels via Liquefaction - Hydrothermal Liquefaction Reactor Design: April 5, 2013

    SciTech Connect

    Knorr, D.; Lukas, J.; Schoen, P.

    2013-11-01

    This report provides detailed reactor designs and capital costs, and operating cost estimates for the hydrothermal liquefaction reactor system, used for biomass-to-biofuels conversion, under development at Pacific Northwest National Laboratory. Five cases were developed and the costs associated with all cases ranged from $22 MM/year - $47 MM/year.

  8. Application of Computational Fluid Dynamics Model to Disinfection Reactors in Water Reclamation Plants

    NASA Astrophysics Data System (ADS)

    Helmns, Andrea; Texeira, Pablo; Issakhanian, Emin; Saez, Jose

    2014-11-01

    California's current drought has renewed public interest in recycled water from Water Reclamation Plants (WRPs). It is critical that the recycled water meets public health standards. This project consists of simulating the transport of an instantaneous conservative tracer through the chlorine contact tanks at two WRPs in California, where recycled water regulations stipulate a minimum 90-minute modal contact time during disinfection at peak dry weather design flow. Computational Fluid Dynamics (CFD) is used to model the turbulent flow, transport, and contact time of a conservative solute for several real operating scenarios. Given as-built drawings and operation parameters, the chlorine contact tanks are modeled to match actual geometries and flow conditions. The turbulent flow solutions are used as the basis to model the transport of a turbulently diffusing conservative tracer added instantaneously to the inlet of the reactors. This tracer simulates the transport through advection and dispersion of chlorine in the WRPs. Breakthrough curves of the tracer at the outlet are used to determine the modal contact times.

  9. The use of LBB concept in French fast reactors: Application to SPX plant

    SciTech Connect

    Turbat, A.; Deschanels, H.; Sperandio, M.

    1997-04-01

    The leak before break (LBB) concept was not used at the design level for SUPERPHENIX (SPX), but different studies have been performed or are in progress concerning different components : Main Vessel (MV), pipings. These studies were undertaken to improve the defense in depth, an approach used in all French reactors. In a first study, the LBB approach has been applied to the MV of SPX plant to verify the absence of risk as regards the core supporting function and to help in the definition of in-service inspection (ISI) program. Defining a reference semi-elliptic defect located in the welds of the structure, it is verified that the crack growth is limited and that the end-of-life defect is smaller than the critical one. Then it is shown that the hoop welds (those which are the most important for safety) located between the roof and the triple point verify the leak-before-break criteria. However, generally speaking, the low level of membrane primary stresses which is favorable for the integrity of the vessel makes the application of the leak-before-break concept more difficult due to small crack opening areas. Finally, the extension of the methodology to the secondary pipings of SPX incorporating recent European works of DCRC is briefly presented.

  10. Conceptual design of a new homogeneous reactor for medical radioisotope Mo-99/Tc-99m production

    SciTech Connect

    Liem, Peng Hong; Tran, Hoai Nam; Sembiring, Tagor Malem; Arbie, Bakri

    2014-09-30

    To partly solve the global and regional shortages of Mo-99 supply, a conceptual design of a nitrate-fuel-solution based homogeneous reactor dedicated for Mo-99/Tc-99m medical radioisotope production is proposed. The modified LEU Cintichem process for Mo-99 extraction which has been licensed and demonstrated commercially for decades by BATAN is taken into account as a key design consideration. The design characteristics and main parameters are identified and the advantageous aspects are shown by comparing with the BATAN's existing Mo-99 supply chain which uses a heterogeneous reactor (RSG GAS multipurpose reactor)

  11. Conceptual design of a new homogeneous reactor for medical radioisotope Mo-99/Tc-99m production

    NASA Astrophysics Data System (ADS)

    Liem, Peng Hong; Tran, Hoai Nam; Sembiring, Tagor Malem; Arbie, Bakri

    2014-09-01

    To partly solve the global and regional shortages of Mo-99 supply, a conceptual design of a nitrate-fuel-solution based homogeneous reactor dedicated for Mo-99/Tc-99m medical radioisotope production is proposed. The modified LEU Cintichem process for Mo-99 extraction which has been licensed and demonstrated commercially for decades by BATAN is taken into account as a key design consideration. The design characteristics and main parameters are identified and the advantageous aspects are shown by comparing with the BATAN's existing Mo-99 supply chain which uses a heterogeneous reactor (RSG GAS multipurpose reactor).

  12. Summary of experimental tests of elastomeric seismic isolation bearings for use in nuclear reactor plants

    SciTech Connect

    Seidensticker, R.W.; Chang, Y.W.; Kulak, R.F.

    1992-01-01

    This paper describes an experimental test program for isolator bearings which was developed to help establish the viability of using laminated elastomer bearings for base isolation of nuclear reactor plants. The goal of the test program is to determine the performance characteristics of laminated seismic isolation bearings under a wide range of loadings. Tests were performed on scale-size laminated seismic isolators both within the design shear strain range to determine the response of the bearing under expected earthquake loading conditions, and beyond the design range to determine failure modes and to establish safety margins. Three types of bearings, each produced from a different manufacturer, have been tested: (1) high shape factor-high damping-high shear modulus bearings; (2) medium shape factor-high damping-high shear modulus bearings; and (3) medium shape factor-high damping-low shear modulus bearings. All of these tests described in this report were performed at the Earthquake Engineering Research Center at the University of California, Berkeley, with technical assistance from ANL. The tests performed on the three types of bearings have confirmed the high performance characteristics of the high damping-high and low shear modulus elastomeric bearings. The bearings have shown that they are capable of having extremely large shear strains before failure occurs. The most common failure mechanism was the debonding of the top steel plate from the isolators. This failure mechanism can be virtually eliminated by improved manufacturing quality control. The most important result of the failure test of the isolators is the fact that bearings can sustain large horizontal displacement, several times larger than the design value, with failure. Their performance in moderate and strong earthquakes will be far superior to conventional structures.

  13. Summary of experimental tests of elastomeric seismic isolation bearings for use in nuclear reactor plants

    SciTech Connect

    Seidensticker, R.W.; Chang, Y.W.; Kulak, R.F.

    1992-05-01

    This paper describes an experimental test program for isolator bearings which was developed to help establish the viability of using laminated elastomer bearings for base isolation of nuclear reactor plants. The goal of the test program is to determine the performance characteristics of laminated seismic isolation bearings under a wide range of loadings. Tests were performed on scale-size laminated seismic isolators both within the design shear strain range to determine the response of the bearing under expected earthquake loading conditions, and beyond the design range to determine failure modes and to establish safety margins. Three types of bearings, each produced from a different manufacturer, have been tested: (1) high shape factor-high damping-high shear modulus bearings; (2) medium shape factor-high damping-high shear modulus bearings; and (3) medium shape factor-high damping-low shear modulus bearings. All of these tests described in this report were performed at the Earthquake Engineering Research Center at the University of California, Berkeley, with technical assistance from ANL. The tests performed on the three types of bearings have confirmed the high performance characteristics of the high damping-high and low shear modulus elastomeric bearings. The bearings have shown that they are capable of having extremely large shear strains before failure occurs. The most common failure mechanism was the debonding of the top steel plate from the isolators. This failure mechanism can be virtually eliminated by improved manufacturing quality control. The most important result of the failure test of the isolators is the fact that bearings can sustain large horizontal displacement, several times larger than the design value, with failure. Their performance in moderate and strong earthquakes will be far superior to conventional structures.

  14. Advanced-power-reactor design concepts and performance characteristics

    NASA Technical Reports Server (NTRS)

    Davison, H. W.; Kirchgessner, T. A.; Springborn, R. H.; Yacobucci, H. G.

    1974-01-01

    Five reactor cooling concepts which allow continued reactor operation following a single rupture of the coolant system are presented for application with the APR. These concepts incorporate convective cooling, double containment, or heat pipes to ensure operation after a coolant line rupture. Based on an evaluation of several control system concepts, a molybdenum clad, beryllium oxide sliding reflector located outside the pressure vessel is recommended.

  15. Design of an Actinide Burning, Lead-Bismuth Cooled Reactor That Produces Low Cost Electricity

    SciTech Connect

    C. Davis; S. Herring; P. MacDonald; K. McCarthy; V. Shah; K. Weaver; J. Buongiorno; R. Ballinger; K. Doyoung; M. Driscoll; P. Hejzler; M. Kazimi; N. Todreas

    1999-07-01

    The purpose of this project is to investigate the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. The choice of lead-bismuth for the reactor coolant is an actinide burning fast reactor offers enhanced safety and reliability. The advantages of lead-bismuth over sodium as a coolant are related to the following material characteristics: chemical inertness with air and water; higher atomic number; lower vapor pressure at operating temperatures; and higher boiling temperature. Given the status of the field, it was agreed that the focus of this investigation in the first two years will be on the assessment of approaches to optimize core and plant arrangements in order to provide maximum safety and economic potential in this type of reactor.

  16. Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

    NASA Astrophysics Data System (ADS)

    Fic, Adam; Składzień, Jan; Gabriel, Michał

    2015-03-01

    Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle), which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle). The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

  17. Safety and core design of large liquid-metal cooled fast breeder reactors

    NASA Astrophysics Data System (ADS)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  18. Design of feedforward controllers for multivariable plants

    NASA Technical Reports Server (NTRS)

    Seraji, H.

    1987-01-01

    Simple methods for the design of feedforward controllers to achieve steady-state disturbance rejection and command tracking in stable multivariable plants are developed in this paper. The controllers are represented by simple and low-order transfer functions and are not based on reconstruction of the states of the commands and disturbances. For unstable plants, it is shown that the present method can be applied directly when an additional feedback controller is employed to stabilize the plant. The feedback and feedforward controllers do not affect each other and can be designed independently based on the open-loop plant to achieve stability, disturbance rejection and command tracking, respectivley. Numerical examples are given for illustration.

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

    SciTech Connect

    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.

  20. Neutronics analyses in support of the conceptual design of the MAPS NTP reactor

    SciTech Connect

    Raepsaet, X.; Lenain, R.

    1996-03-01

    Within the framework of the French nuclear thermal propulsion program called MAPS (Lenain 1996), several neutronics studies and analyses were performed. The aim was to determine the basic design features of a reactor based on the Pebble Bed Reactor concept (Powell 1985) and needing minimum technological developments. In the concern to further enhance the reactor safety posture and to maintain a minimum engine mass breakdown, a beryllium moderated/reflected reactor using highly enriched UO{sub 2} or UC{sub 2} as fuel has been designed with a mean hydrogen core outlet temperature of 2200 K (theoretical ISP of 859 s). The objective of this paper is to give a detailed neutronics analysis of the MAPS reactor. {copyright} {ital 1996 American Institute of Physics.}

  1. Analysis of Reference Design for Nuclear-Assisted Hydrogen Production at 750°C Reactor Outlet Temperature

    SciTech Connect

    Michael G. McKellar; Edwin A. Harvego

    2010-05-01

    The use of High Temperature Electrolysis (HTE) for the efficient production of hydrogen without the greenhouse gas emissions associated with conventional fossil-fuel hydrogen production techniques has been under investigation at the Idaho National Engineering Laboratory (INL) for the last several years. The activities at the INL have included the development, testing and analysis of large numbers of solid oxide electrolysis cells, and the analyses of potential plant designs for large scale production of hydrogen using a high-temperature gas-cooled reactor (HTGR) to provide the process heat and electricity to drive the electrolysis process. The results of this research led to the selection in 2009 of HTE as the preferred concept in the U.S. Department of Energy (DOE) hydrogen technology down-selection process. However, the down-selection process, along with continued technical assessments at the INL, has resulted in a number of proposed modifications and refinements to improve the original INL reference HTE design. These modifications include changes in plant configuration, operating conditions and individual component designs. This report describes the resulting new INL reference design coupled to two alternative HTGR power conversion systems, a Steam Rankine Cycle and a Combined Cycle (a Helium Brayton Cycle with a Steam Rankine Bottoming Cycle). Results of system analyses performed to optimize the design and to determine required plant performance and operating conditions when coupled to the two different power cycles are also presented. A 600 MWt high temperature gas reactor coupled with a Rankine steam power cycle at a thermal efficiency of 44.4% can produce 1.85 kg/s of hydrogen and 14.6 kg/s of oxygen. The same capacity reactor coupled with a combined cycle at a thermal efficiency of 42.5% can produce 1.78 kg/s of hydrogen and 14.0 kg/s of oxygen.

  2. Manufacturing demonstration of microbially mediated zinc sulfide nanoparticles in pilot-plant scale reactors.

    PubMed

    Moon, Ji-Won; Phelps, Tommy J; Fitzgerald, Curtis L; Lind, Randall F; Elkins, James G; Jang, Gyoung Gug; Joshi, Pooran C; Kidder, Michelle; Armstrong, Beth L; Watkins, Thomas R; Ivanov, Ilia N; Graham, David E

    2016-09-01

    The thermophilic anaerobic metal-reducing bacterium Thermoanaerobacter sp. X513 efficiently produces zinc sulfide (ZnS) nanoparticles (NPs) in laboratory-scale (≤ 24-L) reactors. To determine whether this process can be up-scaled and adapted for pilot-plant production while maintaining NP yield and quality, a series of pilot-plant scale experiments were performed using 100-L and 900-L reactors. Pasteurization and N2-sparging replaced autoclaving and boiling for deoxygenating media in the transition from small-scale to pilot plant reactors. Consecutive 100-L batches using new or recycled media produced ZnS NPs with highly reproducible ~2-nm average crystallite size (ACS) and yields of ~0.5 g L(-1), similar to the small-scale batches. The 900-L pilot plant reactor produced ~320 g ZnS without process optimization or replacement of used medium; this quantity would be sufficient to form a ZnS thin film with ~120 nm thickness over 0.5 m width × 13 km length. At all scales, the bacteria produced significant amounts of acetic, lactic, and formic acids, which could be neutralized by the controlled addition of sodium hydroxide without the use of an organic pH buffer, eliminating 98 % of the buffer chemical costs. The final NP products were characterized using XRD, ICP-OES, TEM, FTIR, PL, DLS, HPLC, and C/N analyses, which confirmed that the growth medium without organic buffer enhanced the ZnS NP properties by reducing carbon and nitrogen surface coatings and supporting better dispersivity with similar ACS.

  3. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    SciTech Connect

    Monado, Fiber; Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik; Aziz, Ferhat; Sekimoto, Hiroshi

    2014-02-12

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

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

    SciTech Connect

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

    2002-12-15

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

  5. CONCEPTUAL DESIGN OF A LUNAR REGOLITH CLUSTERED-REACTOR SYSTEM

    SciTech Connect

    John Darrell Bess

    2009-06-01

    It is proposed that a fast-fission, heatpipe-cooled, lunar-surface power reactor system be divided into subcritical units that could be launched safely without the incorporation of additional spectral shift absorbers or other complex means of control. The reactor subunits are to be emplaced directly into the lunar regolith utilizing the regolith not just for shielding but as the reflector material to increase the neutron economy of the system. While a single subunit cannot achieve criticality by itself, coordinated placement of additional subunits will provide a critical reactor system for lunar surface power generation. A lunar regolith clustered-reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of a slight increase in launch mass per rated power level and an overall reduction in neutron economy when compared to a single-reactor system. Additional subunits may be launched with future missions to increase the cluster size and power according to desired lunar base power demand and lifetime. The results address the potential uncertainties associated with the lunar regolith material and emplacement of the subunit systems. Physical distance between subunits within the clustered emplacement exhibits the most significant feedback regarding changes in overall system reactivity. Narrow, deep holes will be the most effective in reducing axial neutron leakage from the core. The variation in iron concentration in the lunar regolith can directly influence the overall system reactivity although its effects are less than the more dominant factors of subunit emplacement.

  6. 78 FR 69139 - Physical Security-Design Certification and Operating Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-18

    ... Nuclear Energy Institute (NEI) submitted a letter on October 9, 2013 (Agencywide Documents Access and... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Physical Security--Design Certification and Operating Reactors AGENCY: Nuclear...

  7. The role of integral experiments and nuclear cross section evaluations in space nuclear reactor design

    NASA Astrophysics Data System (ADS)

    Moses, David L.; McKnight, Richard D.

    The importance of the nuclear and neutronic properties of candidate space reactor materials to the design process has been acknowledged as has been the use of benchmark reactor physics experiments to verify and qualify analytical tools used in design, safety, and performance evaluation. Since June 1966, the Cross Section Evaluation Working Group (CSEWG) has acted as an interagency forum for the assessment and evaluation of nuclear reaction data used in the nuclear design process. CSEWG data testing has involved the specification and calculation of benchmark experiments which are used widely for commercial reactor design and safety analysis. These benchmark experiments preceded the issuance of the industry standards for acceptance, but the benchmarks exceed the minimum acceptance criteria for such data. Thus, a starting place has been provided in assuring the accuracy and uncertainty of nuclear data important to space reactor applications.

  8. Ammonia plant designers talk of big energy savings

    SciTech Connect

    Axelrod, L.C.

    1980-08-27

    The ammonia plant that Pullman Kellogg has designed for Sherritt-Gordon Mines Ltd. in Alberta will require < 27 million Btu/ton of ammonia, and save $8.10/ton in energy costs because of improvements involving increased pressure in the primary reformer; more efficient use of the heat from the secondary reformer; carbon dioxide recovery by Allied Chemical Corp.'s Selexol process; the reduction of power requirements in the synthesis recycle loop; and the use of a horizontal reactor. C. F. Braun and Co. claims that its Purifier process will require < 25 million Btu/ton, due to the use of excess air in the secondary reformer. C-E Lummus offers a 1500 ton/day plant which, incorporating cryogenic recovery of hydrogen from purge gas and operation at a lower steam-to-carbon ratio, would require only 26 million Btu/ton; Haldor Topsoe Inc. offers a design rated at 26.3 million Btu/ton. According to L. C. Axelrod of Pullman Kellogg, ammonia plant construction will shift to gas-rich areas outside the U.S. and Europe. The 3% of the U.S. natural gas used by the fertilizer industry accounts for > 95% of ammonia feedstock.

  9. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  10. Design change management in regulation of nuclear fleets: World nuclear association's working groups on Cooperation in Reactor Design Evaluation and Licensing (CORDEL)

    SciTech Connect

    Swinburn, R.; Borysova, I.; Waddington, J.; Head, J. G.; Raidis, Z.

    2012-07-01

    The 60 year life of a reactor means that a plant will undergo change during its life. To ensure continuing safety, changes must be made with a full understanding of the design intent. With this aim, regulators require that each operating organisation should have a formally designated entity responsible for complete design knowledge in regard to plant safety. INSAG-19 calls such an entity 'Design Authority'. This requirement is difficult to achieve, especially as the number of countries and utilities operating plants increases. Some of these operating organisations will be new, and some will be small. For Gen III plants sold on a turnkey basis, it is even more challenging for the operating company to develop and retain the full knowledge needed for this role. CORDEL's Task Force entitled 'Design Change Management' is investigating options for effective design change management with the aim to support design standardization throughout a fleet's lifetime by means of enhanced international cooperation within industry and regulators. This paper starts with considering the causes of design change and identifies reasons for the increased beneficial involvement of the plant's original vendor in the design change process. A key central theme running through the paper is the definition of responsibilities for design change. Various existing mechanisms of vendor-operator interfaces over design change and how they are managed in different organisational and regulatory environments around the world are considered, with the functionality of Owners Groups and Design Authority being central. The roles played in the design change process by vendors, utilities, regulators, owners' groups and other organisations such as WANO are considered The aerospace industry approach to Design Authority has been assessed to consider what lessons might be learned. (authors)

  11. Fluid Flow Characteristic Simulation of the Original TRIGA 2000 Reactor Design Using Computational Fluid Dynamics Code

    SciTech Connect

    Fiantini, Rosalina; Umar, Efrizon

    2010-06-22

    Common energy crisis has modified the national energy policy which is in the beginning based on natural resources becoming based on technology, therefore the capability to understanding the basic and applied science is needed to supporting those policies. National energy policy which aims at new energy exploitation, such as nuclear energy is including many efforts to increase the safety reactor core condition and optimize the related aspects and the ability to build new research reactor with properly design. The previous analysis of the modification TRIGA 2000 Reactor design indicates that forced convection of the primary coolant system put on an effect to the flow characteristic in the reactor core, but relatively insignificant effect to the flow velocity in the reactor core. In this analysis, the lid of reactor core is closed. However the forced convection effect is still presented. This analysis shows the fluid flow velocity vector in the model area without exception. Result of this analysis indicates that in the original design of TRIGA 2000 reactor, there is still forced convection effects occur but less than in the modified TRIGA 2000 design.

  12. Design and Testing of a Boron Carbide Capsule for Spectral Tailoring in Mixed-Spectrum Reactors

    SciTech Connect

    Greenwood, Lawrence R.; Wittman, Richard S.; Pierson, Bruce D.; Metz, Lori A.; Payne, Rosara F.; Finn, Erin C.; Friese, Judah I.

    2012-03-01

    A boron carbide capsule has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State University. Irradiations were conducted in pulsed mode and in continuous operation for up to 4 hours. A cadmium cover was used to reduce thermal heating. The neutron spectrum calculated with MCNP was found to be in good agreement with reactor dosimetry measurements using the STAY'SL computer code. The neutron spectrum resembles that of a fast reactor. Design of a capsule using boron carbide enriched in {sup 10}B shows that it is possible to produce a neutron spectrum similar to {sup 235}U fission.

  13. CALIOP: a multichannel design code for gas-cooled fast reactors. Code description and user's guide

    SciTech Connect

    Thompson, W.I.

    1980-10-01

    CALIOP is a design code for fluid-cooled reactors composed of parallel fuel tubes in hexagonal or cylindrical ducts. It may be used with gaseous or liquid coolants. It has been used chiefly for design of a helium-cooled fast breeder reactor and has built-in cross section information to permit calculations of fuel loading, breeding ratio, and doubling time. Optional cross-section input allows the code to be used with moderated cores and with other fuels.

  14. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  15. Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report

    SciTech Connect

    2000-08-01

    OAK B188 Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report. The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-formed approach for the design and regulation of nuclear power plants. This approach will include the development and/or confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRS) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go further by focusing on the design of new plants.

  16. Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

    DOE PAGES

    Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.; ...

    2016-12-21

    Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate

  17. Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

    SciTech Connect

    Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.; Carbajo, Juan J.; Greenwood, M. Scott; Hale, Richard; Harrison, Thomas J.; Powers, Jeffrey J.; Robb, Kevin R.; Terrell, Jerry; Wysocki, Aaron J.; Gehin, Jess C.; Worrall, Andrew

    2016-12-21

    Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It

  18. Manufacturing demonstration of microbially mediated zinc sulfide nanoparticles in pilot-plant scale reactors

    SciTech Connect

    Moon, Ji-Won; Phelps, Tommy J.; Fitzgerald Jr, Curtis L.; Lind, Randall F.; Elkins, James G.; Jang, Gyoung Gug; Joshi, Pooran C.; Kidder, Michelle; Armstrong, Beth L.; Watkins, Thomas R.; Ivanov, Ilia N.; Graham, David E.

    2016-04-27

    The thermophilic anaerobic metal-reducing bacterium Thermoanaerobacter sp. X513 efficiently produces zinc sulfide (ZnS) nanoparticles (NPs) in laboratory-scale ( ≤24-L) reactors. To determine whether this process can be up-scaled and adapted for pilot-plant production while maintaining NP yield and quality, a series of meso-scale experiments were performed using 100-l and 900-l reactors. Pasteurization and N2-sparging replaced autoclaving and boiling for deoxygenating media in the transition from small-scale to pilot-plant reactors. Consecutive 100-L batches using new or recycled media produced ZnS NPs with highly reproducible ~2 nm average crystallite size (ACS) and yields of ~0.5g L-1, similar to small-scale batches. The 900-L pilot plant reactor produced ~ 320 g ZnS without process optimization or replacement of used medium; this quantity would be sufficient to form a ZnS thin film with ~120 nm thickness over 0.5 m width 13 km length. At all scales, the bacteria produced significant amounts of acetic, lactic and formic acids, which could be neutralized by the controlled addition of sodium hydroxide without the use of an organic pH buffer, eliminating 98% of the buffer chemical costs. In conclusion, the final NP products were characterized using XRD, ICP-OES, FTIR, DLS, and C/N analyses, which confirmed the growth medium without organic buffer enhanced the ZnS NP properties by reducing carbon and nitrogen surface coatings and supporting better dispersivity with similar ACS.

  19. Manufacturing demonstration of microbially mediated zinc sulfide nanoparticles in pilot-plant scale reactors

    DOE PAGES

    Moon, Ji-Won; Phelps, Tommy J.; Fitzgerald Jr, Curtis L.; ...

    2016-04-27

    The thermophilic anaerobic metal-reducing bacterium Thermoanaerobacter sp. X513 efficiently produces zinc sulfide (ZnS) nanoparticles (NPs) in laboratory-scale ( ≤24-L) reactors. To determine whether this process can be up-scaled and adapted for pilot-plant production while maintaining NP yield and quality, a series of meso-scale experiments were performed using 100-l and 900-l reactors. Pasteurization and N2-sparging replaced autoclaving and boiling for deoxygenating media in the transition from small-scale to pilot-plant reactors. Consecutive 100-L batches using new or recycled media produced ZnS NPs with highly reproducible ~2 nm average crystallite size (ACS) and yields of ~0.5g L-1, similar to small-scale batches. The 900-Lmore » pilot plant reactor produced ~ 320 g ZnS without process optimization or replacement of used medium; this quantity would be sufficient to form a ZnS thin film with ~120 nm thickness over 0.5 m width 13 km length. At all scales, the bacteria produced significant amounts of acetic, lactic and formic acids, which could be neutralized by the controlled addition of sodium hydroxide without the use of an organic pH buffer, eliminating 98% of the buffer chemical costs. In conclusion, the final NP products were characterized using XRD, ICP-OES, FTIR, DLS, and C/N analyses, which confirmed the growth medium without organic buffer enhanced the ZnS NP properties by reducing carbon and nitrogen surface coatings and supporting better dispersivity with similar ACS.« less

  20. Development of a neutronics calculation method for designing commercial type Japanese sodium-cooled fast reactor

    SciTech Connect

    Takeda, T.; Shimazu, Y.; Hibi, K.; Fujimura, K.

    2012-07-01

    Under the R and D project to improve the modeling accuracy for the design of fast breeder reactors the authors are developing a neutronics calculation method for designing a large commercial type sodium- cooled fast reactor. The calculation method is established by taking into account the special features of the reactor such as the use of annular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verification and Validation, and Uncertainty Qualification (V and V and UQ) of the calculation method is being performed by using measured data from the prototype FBR Monju. The results of this project will be used in the design and analysis of the commercial type demonstration FBR, known as the Japanese Sodium fast Reactor (JSFR). (authors)

  1. Microwave radiation and reactor design influence microbial communities during methane fermentation.

    PubMed

    Cydzik-Kwiatkowska, Agnieszka; Zieliński, Marcin; Jaranowska, Paulina

    2012-09-01

    The effect of reactor design and method of heating on the efficiency of methane fermentation and composition of microbial communities, especially methanogenic Archaea, were determined. The research was carried out using submerge- and trickling-bed reactors fed with wastewater and the heat supply into the reactors included a convection heating method and microwave radiation. The polymerase chain reaction-denaturing gradient gel electrophoresis and relative real-time PCR were used in order to assess the biofilm communities. The best fermentation results and the highest abundance of methanogenic Archaea in biomass were observed in microwave heated trickling-bed reactors. The research proved that in reactors of identical design, the application of microwaves enabled a higher fermentation efficiency to be obtained and simultaneously increased the diversity of methanogenic Archaea communities that favors process stability. All the identified sequences of Archaea belonged to Methanosarcina sp., suggesting that species from this genera are susceptible to non-thermal effects of microwaves. There were no effects from microwaves on the bacterial communities in both types of reactors, however, the bacterial species composition varied in the reactors of different design.

  2. Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H. Volume 2, Final report

    SciTech Connect

    Waganer, L.M.; Driemeyer, D.E.; Lee, V.D.

    1992-03-01

    This report contains a review of design studies for Inertial Confinement reactor. This second of three volumes discussions is some detail the following: Objectives, requirements, and assumptions; rationale for design option selection; key technical issues and R&D requirements; and conceptual design selection and description.

  3. 75 FR 78777 - Advisory Committee On Reactor Safeguards; Renewal

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-12-16

    ...: The Advisory Committee on Reactor Safeguards was established by Section 29 of the Atomic Energy Act... accident phenomena; design of nuclear power plant structures, systems and components; materials...

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

    SciTech Connect

    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.

  5. Design related aspects in advanced nuclear fission plants

    NASA Astrophysics Data System (ADS)

    Hoffelner, Wolfgang

    2011-02-01

    Important issues to be considered for design of future reactors are: extrapolation of stress rupture data, creep-fatigue, negligible creep, damage monitoring. The paper highlights some new developments taking examples from a martensitic steel (mod 9% Cr), oxide dispersion strengthened (ODS) steels and nickel-base superalloys. Traditional approaches to extrapolation of (thermal) stress rupture data like Larson-Miller Parameter or Monkman-Grant rule seem to be valid concepts also for advanced reactors. However, a significant influence of cyclic softening on creep rates and stress rupture data can be expected as shown for grade 91. This is particularly true for creep-fatigue interactions. Based on cyclic stress-strain behaviour it is also possible to get very good life-time predictions under creep-fatigue with a strain range separation (inelastic fatigue and creep ranges) technique which could replace the currently used linear life fraction rule. Results from in-beam irradiation creep reveal no significant influence of dispersoid size. It can be assumed that irradiation creep is a matrix property. Finally it is shown that micro-sample testing of exposed material could be used as an advanced method for damage assessment in future nuclear power plants.

  6. Seismic design technology for Breeder Reactor structures. Volume 3: special topics in reactor structures

    SciTech Connect

    Reddy, D.P.

    1983-04-01

    This volume is divided into six chapters: analysis techniques, equivalent damping values, probabilistic design factors, design verifications, equivalent response cycles for fatigue analysis, and seismic isolation. (JDB)

  7. Assessment of next generation nuclear plant intermediate heat exchanger design.

    SciTech Connect

    Majumdar, S.; Moisseytsev, A.; Natesan, K.; Nuclear Engineering Division

    2008-10-17

    The Next Generation Nuclear Plant (NGNP), which is an advanced high temperature gas reactor (HTGR) concept with emphasis on production of both electricity and hydrogen, involves helium as the coolant and a closed-cycle gas turbine for power generation with a core outlet/gas turbine inlet temperature of 900-1000 C. In the indirect cycle system, an intermediate heat exchanger is used to transfer the heat from primary helium from the core to the secondary fluid, which can be helium, nitrogen/helium mixture, or a molten salt. The system concept for the vary high temperature reactor (VHTR) can be a reactor based on the prismatic block of the GT-MHR developed by a consortium led by General Atomics in the U.S. or based on the PBMR design developed by ESKOM of South Africa and British Nuclear Fuels of U.K. This report has made an assessment on the issues pertaining to the intermediate heat exchanger (IHX) for the NGNP. A detailed thermal hydraulic analysis, using models developed at ANL, was performed to calculate heat transfer, temperature distribution, and pressure drop. Two IHX designs namely, shell and straight tube and compact heat exchangers were considered in an earlier assessment. Helical coil heat exchangers were analyzed in the current report and the results were compared with the performance features of designs from industry. In addition, a comparative analysis is presented between the shell and straight tube, helical, and printed circuit heat exchangers from the standpoint of heat exchanger volume, primary and secondary sides pressure drop, and number of tubes. The IHX being a high temperature component, probably needs to be designed using ASME Code Section III, Subsection NH, assuming that the IHX will be classified as a class 1 component. With input from thermal hydraulic calculations performed at ANL, thermal conduction and stress analyses were performed for the helical heat exchanger design and the results were compared with earlier-developed results on

  8. Seismic Analysis Issues in Design Certification Applications for New Reactors

    SciTech Connect

    Miranda, M.; Morante, R.; Xu, J.

    2011-07-17

    The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early reso- lution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific de- sign parameters are within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of several seismic analysis issues encountered during a review of recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.

  9. Innovative Design of New Geothermal Generating Plants

    SciTech Connect

    Bloomquist, R. Gordon; Geyer, John D.; Sifford, B. Alexander III

    1989-07-01

    This very significant and useful report assessed state-of-the-art geothermal technologies. The findings presented in this report are the result of site visits and interviews with plant owners and operators, representatives of major financial institutions, utilities involved with geothermal power purchases and/or wheeling. Information so obtained was supported by literature research and data supplied by engineering firms who have been involved with designing and/or construction of a majority of the plants visited. The interviews were conducted by representatives of the Bonneville Power Administration, the Washington State Energy Office, and the Oregon Department of Energy during the period 1986-1989. [DJE-2005

  10. A full scale worm reactor for efficient sludge reduction by predation in a wastewater treatment plant.

    PubMed

    Tamis, J; van Schouwenburg, G; Kleerebezem, R; van Loosdrecht, M C M

    2011-11-15

    Sludge predation can be an effective solution to reduce sludge production at a wastewater treatment plant. Oligochaete worms are the natural consumers of biomass in benthic layers in ecosystems. In this study the results of secondary sludge degradation by the aquatic Oligochaete worm Aulophorus furcatus in a 125 m(3) reactor and further sludge conversion in an anaerobic tank are presented. The system was operated over a period of 4 years at WWTP Wolvega, the Netherlands and was fed with secondary sludge from a low loaded activated sludge process. It was possible to maintain a stable and active population of the aquatic worm species A. furcatus during the full period. Under optimal conditions a sludge conversion of 150-200 kg TSS/d or 1.2-1.6 kg TSS/m(3)/d was established in the worm reactor. The worms grew as a biofilm on carrier material in the reactor. The surface specific conversion rate reached 140-180 g TSS/m(2)d and the worm biomass specific conversion rate was 0.5-1 g TSS sludge/g dry weight worms per day. The sludge reduction under optimal conditions in the worm reactor was 30-40%. The degradation by worms was an order of magnitude larger than the endogenous conversion rate of the secondary sludge. Effluent sludge from the worm reactor was stored in an anaerobic tank where methanogenic processes became apparent. It appeared that besides reducing the sludge amount, the worms' activity increased anaerobic digestibility, allowing for future optimisation of the total system by maximising sludge reduction and methane formation. In the whole system it was possible to reduce the amount of sludge by at least 65% on TSS basis. This is a much better total conversion than reported for anaerobic biodegradability of secondary sludge of 20-30% efficiency in terms of TSS reduction.

  11. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    SciTech Connect

    Schlereth, J.R.; Pennington, D.

    1996-12-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it`s Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components.

  12. Conceptual design of fuel transfer cask for Reactor TRIGA PUSPATI (RTP)

    NASA Astrophysics Data System (ADS)

    Muhamad, Shalina Sheik; Hamzah, Mohd Arif Arif B.

    2014-02-01

    Spent fuel transfer cask is used to transfer a spent fuel from the reactor tank to the spent fuel storage or for spent fuel inspection. Typically, the cask made from steel cylinders that are either welded or bolted closed. The cylinder is enclosed with additional steel, concrete, or other material to provide radiation shielding and containment of the spent fuel. This paper will discuss the Conceptual Design of fuel transfer cask for Reactor TRIGA Puspati (RTP).

  13. Design of unique pins for irradiation of higher actinides in a fast reactor

    SciTech Connect

    Basmajian, J.A.; Birney, K.R.; Weber, E.T.; Adair, H.L.; Quinby, T.C.; Raman, S.; Butler, J.K.; Bateman, B.C.; Swanson, K.M.

    1982-03-01

    The actinides produced by transmutation reactions in nuclear reactor fuels are a significant factor in nuclear fuel burnup, transportation and reprocessing. Irradiation testing is a primary source of data of this type. A segmented pin design was developed which provides for incorporation of multiple specimens of actinide oxides for irradiation in the UK's Prototype Fast Reactor (PFR) at Dounreay Scotland. Results from irradiation of these pins will extend the basic neutronic and material irradiation behavior data for key actinide isotopes.

  14. Reactor Design and Decommissioning - An Overview of International Activities in Post Fukushima Era1 - 12396

    SciTech Connect

    Devgun, Jas S.; Laraia, Michele; Dinner, Paul

    2012-07-01

    Accidents at the Fukushima Dai-ichi reactors as a result of the devastating earthquake and tsunami of March 11, 2011 have not only dampened the nuclear renaissance but have also initiated a re-examination of the design and safety features for the existing and planned nuclear reactors. Even though failures of some of the key site features at Fukushima can be attributed to events that in the past would have been considered as beyond the design basis, the industry as well as the regulatory authorities are analyzing what features, especially passive features, should be designed into the new reactor designs to minimize the potential for catastrophic failures. It is also recognized that since the design of the Fukushima BWR reactors which were commissioned in 1971, many advanced safety features are now a part of the newer reactor designs. As the recovery efforts at the Fukushima site are still underway, decisions with respect to the dismantlement and decommissioning of the damaged reactors and structures have not yet been finalized. As it was with Three Mile Island, it could take several decades for dismantlement, decommissioning and clean up, and the project poses especially tough challenges. Near-term assessments have been issued by several organizations, including the IAEA, the USNRC and others. Results of such investigations will lead to additional improvements in system and site design measures including strengthening of the anti-tsunami defenses, more defense-in-depth features in reactor design, and better response planning and preparation involving reactor sites. The question also arises what would the effect be on the decommissioning scene worldwide, and what would the effect be on the new reactors when they are eventually retired and dismantled. This paper provides an overview of the US and international activities related to recovery and decommissioning including the decommissioning features in the reactor design process and examines these from a new

  15. The impact of thermal fatigue and other design limits on pulsed commercial tokamak reactor design

    NASA Astrophysics Data System (ADS)

    LeClaire, R. J.; Meyer, J. E.

    1985-10-01

    Methods of analysis for fusion first-wall design are developed. Several design limits have been evaluated and combined to present tradeoffs in the form of design windows. These considerations include limits related to thermal fatigue, primary membrane strength, displacement under loading, ratcheting, radiation damage, and plasma-wall interactions. Special emphasis is placed on the investigation of thermal fatigue using a two-dimensional treatment of a tubular first-wall configuration. The work is motivated by the proposal of the Ultra Long Pulse Commercial Reactor (ULTR), a machine capable of delivering plasma burn pulses of up to 24 hr in length. The present work looks in detail at the impact of pertinent characteristics of the first-wall design, such as pulse length, coolant pressure, first-wall thickness, and first-wall lifetime on the structural effects considered. Computer programs are developed and are used to consider several major structural effects on a cylindrical first-wall element for both 316 stainless steel and vanadium alloy. Results indicate that short pulse lengths (greater than a few minutes) can be tolerated in tokamak operation. For stainless steel this is true for heat depositions up to 1 MW/m2, while vanadium can tolerate heat depositions as high as 2 MW/m2. Long pulse operation can be used to increase modestly the allowable heat deposition or to increase useful wall thickness by 1-2 mm. It appears that irradiation swelling and embrittlement, not fatigue, ultimately limits the first-wall design.

  16. The Caltech Photooxidation Flow Tube reactor: design, fluid dynamics and characterization

    NASA Astrophysics Data System (ADS)

    Huang, Yuanlong; Coggon, Matthew M.; Zhao, Ran; Lignell, Hanna; Bauer, Michael U.; Flagan, Richard C.; Seinfeld, John H.

    2017-03-01

    Flow tube reactors are widely employed to study gas-phase atmospheric chemistry and secondary organic aerosol (SOA) formation. The development of a new laminar-flow tube reactor, the Caltech Photooxidation Flow Tube (CPOT), intended for the study of gas-phase atmospheric chemistry and SOA formation, is reported here. The present work addresses the reactor design based on fluid dynamical characterization and the fundamental behavior of vapor molecules and particles in the reactor. The design of the inlet to the reactor, based on computational fluid dynamics (CFD) simulations, comprises a static mixer and a conical diffuser to facilitate development of a characteristic laminar flow profile. To assess the extent to which the actual performance adheres to the theoretical CFD model, residence time distribution (RTD) experiments are reported with vapor molecules (O3) and submicrometer ammonium sulfate particles. As confirmed by the CFD prediction, the presence of a slight deviation from strictly isothermal conditions leads to secondary flows in the reactor that produce deviations from the ideal parabolic laminar flow. The characterization experiments, in conjunction with theory, provide a basis for interpretation of atmospheric chemistry and SOA studies to follow. A 1-D photochemical model within an axially dispersed plug flow reactor (AD-PFR) framework is formulated to evaluate the oxidation level in the reactor. The simulation indicates that the OH concentration is uniform along the reactor, and an OH exposure (OHexp) ranging from ˜ 109 to ˜ 1012 molecules cm-3 s can be achieved from photolysis of H2O2. A method to calculate OHexp with a consideration for the axial dispersion in the present photochemical system is developed.

  17. Design of an Actinide-Burning, Lead or Lead-Bismuth Cooled Reactor that Produces Low-Cost Electricity

    SciTech Connect

    Mac Donald, Philip Elsworth; Weaver, Kevan Dean; Davis, Cliff Bybee; MIT folks

    2000-07-01

    The purpose of this Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology (MIT) University Research Consortium (URC) project is to investigate the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. Work has been accomplished in four major areas of research: core neutronic design, material compatibility, plant engineering, and coolant activation. In the area of core neutronic design, the reactivity vs. burnup and discharge isotopics of both non-fertile and fertile fuels were evaluated. An innovative core for pure actinide burning that uses streaming, fertile-free fuel assemblies was studied in depth. This particular core exhibits excellent reactivity performance upon coolant voiding, even for voids that occur in the core center, and has a transuranic (TRU) destruction rate that is comparable to the proposed accelerator transmutation of waste (ATW) facility. These studies suggest that a core can be designed to achieve a long life while maintaining safety and minimizing waste. In the area of material compatibility studies, an experimental apparatus for the investigation of the flow-assisted dissolution and precipitation (corrosion) of potential fuel cladding and structural materials has been designed and built at the INEEL. The INEEL forced-convection corrosion cell consists of a small heated vessel with a shroud and gas flow system. The corrosion cell is being used to test steel that is commercially available in the United States to temperatures above 650°C. Progress in plant engineering was made for two reactor concepts, one utilizing an indirect cycle with heat exchangers and the other utilizing a direct-contact steam cycle. The evaluation of the

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

    SciTech Connect

    Not Available

    1986-10-01

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

  19. DU-AGG pilot plant design study

    SciTech Connect

    Lessing, P.A.; Gillman, H.

    1996-07-01

    The Idaho National Engineering Laboratory (INEL) is developing new methods to produce high-density aggregate (artificial rock) primarily consisting of depleted uranium oxide. The objective is to develop a low-cost method whereby uranium oxide powder (UO[sub 2], U[sub 3]O[sub ]8, or UO[sub 3]) can be processed to produce high-density aggregate pieces (DU-AGG) having physical properties suitable for disposal in low-level radioactive disposal facilities or for use as a component of high-density concrete used as shielding for radioactive materials. A commercial company, G-M Systems, conducted a design study for a manufacturing pilot plant to process DU-AGG. The results of that study are included and summarized in this report. Also explained are design considerations, equipment capacities, the equipment list, system operation, layout of equipment in the plant, cost estimates, and the proposed plan and schedule.

  20. Growth of plant root cultures in liquid- and gas-dispersed reactor environments.

    PubMed

    McKelvey, S A; Gehrig, J A; Hollar, K A; Curtis, W R

    1993-01-01

    The growth of Agrobacterium transformed "hairy root" cultures of Hyoscyamus muticus was examined in various liquid- and gas-dispersed bioreactor configurations. Reactor runs were replicated to provide statistical comparisons of nutrient availability on culture performance. Accumulated tissue mass in submerged air-sparged reactors was 31% of gyratory shake-flask controls. Experiments demonstrate that poor performance of sparged reactors is not due to bubble shear damage, carbon dioxide stripping, settling, or flotation of roots. Impaired oxygen transfer due to channeling and stagnation of the liquid phase are the apparent causes of poor growth. Roots grown on a medium-perfused inclined plane grew at 48% of gyratory controls. This demonstrates the ability of cultures to partially compensate for poor liquid distribution through vascular transport of nutrients. A reactor configuration in which the medium is sprayed over the roots and permitted to drain down through the root tissue was able to provide growth rates which are statistically indistinguishable (95% T-test) from gyratory shake-flask controls. In this type of spray/trickle-bed configuration, it is shown that distribution of the roots becomes a key factor in controlling the rate of growth. Implications of these results regarding design and scale-up of bioreactors to produce fine chemicals from root cultures are discussed.

  1. Design and construction of a cascading pressure reactor prototype for solar-thermochemical hydrogen production

    NASA Astrophysics Data System (ADS)

    Ermanoski, Ivan; Grobbel, Johannes; Singh, Abhishek; Lapp, Justin; Brendelberger, Stefan; Roeb, Martin; Sattler, Christian; Whaley, Josh; McDaniel, Anthony; Siegel, Nathan P.

    2016-05-01

    Recent work regarding the efficiency maximization for solar thermochemical fuel production in two step cycles has led to the design of a new type of reactor—the cascading pressure reactor—in which the thermal reduction step of the cycle is completed in multiple stages, at successively lower pressures. This approach enables lower thermal reduction pressures than in single-staged reactors, and decreases required pump work, leading to increased solar to fuel efficiencies. Here we report on the design and construction of a prototype cascading pressure reactor and testing of some of the key components. We especially focus on the technical challenges particular to the design, and their solutions.

  2. Licensing topical report: interpretation of general design criteria for high-temperature gas-cooled reactors

    SciTech Connect

    Orvis, D.D.; Raabe, P.H.

    1980-01-01

    This Licensing Topical Report presents a set of General Design Criteria (GDC) which is proposed for applicability to licensing of graphite-moderated, high-temperature gas-cooled reactors (HTGRs). Modifications as necessary to reflect HTGR characteristics and design practices have been made to the GDC derived for applicability to light-water-cooled reactors and presented in Appendix A of Part 50, Title 10, Code of Federal Regulations, including the Introduction, Definitions, and Criteria. It is concluded that the proposed set of GDC affords a better basis for design and licensing of HTGRs.

  3. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    SciTech Connect

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  4. How should future reactors de designated or characterized

    SciTech Connect

    Pedersen, T.; Kupitz, J.

    1990-06-01

    The term inherent safety, or inherently safe, has become widely used and misused throughout the world over the last decade. This document presents a consensus on the meaning and use of a number of descriptive terms for so-called passively safe reactors. The overall purpose of this effort was to start the process of describing and understanding such terms. Eliminating current confusion and misuse of terms by members of the nuclear community would render the terms more meaningful and improve communication. Precisely described technical meanings, corresponding to common usage, to terms used in public discourse, and in other technologies should enhance the credibility of the community with the public.

  5. Feedback system design with an uncertain plant

    NASA Technical Reports Server (NTRS)

    Milich, D.; Valavani, L.; Athans, M.

    1986-01-01

    A method is developed to design a fixed-parameter compensator for a linear, time-invariant, SISO (single-input single-output) plant model characterized by significant structured, as well as unstructured, uncertainty. The controller minimizes the H(infinity) norm of the worst-case sensitivity function over the operating band and the resulting feedback system exhibits robust stability and robust performance. It is conjectured that such a robust nonadaptive control design technique can be used on-line in an adaptive control system.

  6. Feedback system design with an uncertain plant

    NASA Technical Reports Server (NTRS)

    Milich, D.; Valavani, L.; Athans, M.

    1986-01-01

    A method is developed to design a fixed-parameter compensator for a linear, time-invariant, SISO (single-output single-output) plant model characterized by significant structured, as well as unstructured, uncertainty. The controller minimizes the H(infinity) norm of the worst-case sensitivity function over the operating band and the resulting feedback system exhibits robust stability and robust performance. It is conjectured that such a robust nonadaptive control design technique can be used on-line in an adaptive control system.

  7. Assessments of Longevity of Equipment Metal of Nuclear Power Plants equipped with Reactors VVER-1000

    SciTech Connect

    Gorbatykh, V.P.; Al Kassem, S.N.

    2004-07-01

    Characteristics of damage processes of metal of coffer-dams of steam generators collectors at nuclear power plants (NPPs) equipped with reactors VVER-1000 have been mentioned; principles of construction of longevity function has been cited and new approach has been shown while solving the problem of the longevity of the metal resource by substantiating the technological actions with new mode characteristics, performed with the help of specially developed equations and formulae, where practically all damage processes and all influencing factors can be accounted. (authors)

  8. Complete study of the pyrolysis and gasification of scrap tires in a pilot plant reactor.

    PubMed

    Conesa, Juan A; Martín-Gullón, I; Font, R; Jauhiainen, J

    2004-06-01

    The pyrolysis and gasification of tires was studied in a pilot plant reactor provided with a system for condensation of semivolatile matter. The study comprises experiments at 450, 750, and 1000 degrees C both in nitrogen and 10% oxygen atmospheres. Analysis of all the products obtained (gases, liquids, char, and soot) are presented. In the gas phase only methane and benzene yields increase with temperature until 1000 degrees C. In the liquids the main components are styrene, limonene, and isoprene. The solid fraction (including soot) increases with temperature. Zinc content of the char decreases with increasing temperature.

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

    SciTech Connect

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

    2015-03-01

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

  10. Next Generation Nuclear Plant (NGNP) Prismatic HTGR Conceptual Design Project - Final Technical Report

    SciTech Connect

    Saurwein, John

    2011-07-15

    This report is the Final Technical Report for the Next Generation Nuclear Plant (NGNP) Prismatic HTGR Conceptual Design Project conducted by a team led by General Atomics under DOE Award DE-NE0000245. The primary overall objective of the project was to develop and document a conceptual design for the Steam Cycle Modular Helium Reactor (SC-MHR), which is the reactor concept proposed by General Atomics for the NGNP Demonstration Plant. The report summarizes the project activities over the entire funding period, compares the accomplishments with the goals and objectives of the project, and discusses the benefits of the work. The report provides complete listings of the products developed under the award and the key documents delivered to the DOE.

  11. NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) PROGRAM GRANT NUMBER DE-FG03-00SF22168 TECHNICAL PROGRESS REPORT (Nov. 15, 2001 - Feb. 15,2002) ''Design and Layout Concepts for Compact, Factory-Produced, Transportable, Generation IV Reactor Systems''

    SciTech Connect

    Fred R. Mynatt; Andy Kadak; Marc Berte; Larry Miller; Mohammed Khan; Joe McConn; Lawrence Townsend; Wesley Williams; Martin Williamson

    2002-03-15

    The objectives of this project are to develop and evaluate nuclear power plant designs and layout concepts to maximize the benefits of compact modular Generation IV reactor concepts including factory fabrication and packaging for optimal transportation and siting. Three nuclear power plant concepts are being studied representing water, helium and lead-bismuth coolants. This is the sixth quarterly progress report.

  12. Thermal and neutron-physical features of the nuclear reactor for a power pulsation plant for space applications

    NASA Astrophysics Data System (ADS)

    Gordeev, É. G.; Kaminskii, A. S.; Konyukhov, G. V.; Pavshuk, V. A.; Turbina, T. A.

    2012-05-01

    We have explored the possibility of creating small-size reactors with a high power output with the provision of thermal stability and nuclear safety under standard operating conditions and in emergency situations. The neutron-physical features of such a reactor have been considered and variants of its designs preserving the main principles and approaches of nuclear rocket engine technology are presented.

  13. Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing 235U, 233U, and 232Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

    NASA Astrophysics Data System (ADS)

    Ioffe, B. L.; Kochurov, B. P.

    2012-02-01

    A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of 235U. It operates in the open-cycle mode involving 233U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

  14. Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing {sup 235}U, {sup 233}U, and {sup 232}Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

    SciTech Connect

    Ioffe, B. L.; Kochurov, B. P.

    2012-02-15

    A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of {sup 235}U. It operates in the open-cycle mode involving {sup 233}U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

  15. A Virtual Reality Framework to Optimize Design, Operation and Refueling of GEN-IV Reactors.

    SciTech Connect

    Rizwan-uddin; Nick Karancevic; Stefano Markidis; Joel Dixon; Cheng Luo; Jared Reynolds

    2008-04-23

    many GEN-IV candidate designs are currently under investigation. Technical issues related to material, safety and economics are being addressed at research laboratories, industry and in academia. After safety, economic feasibility is likely to be the most important crterion in the success of GEN-IV design(s). Lessons learned from the designers and operators of GEN-II (and GEN-III) reactors must play a vital role in achieving both safety and economic feasibility goals.

  16. Design of the DEMO Fusion Reactor Following ITER

    PubMed Central

    Garabedian, Paul R.; McFadden, Geoffrey B.

    2009-01-01

    Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task. PMID:27504224

  17. Supervisory control design based on hybrid systems and fuzzy events detection. Application to an oxichlorination reactor.

    PubMed

    Altamiranda, Edmary; Torres, Horacio; Colina, Eliezer; Chacón, Edgar

    2002-10-01

    This paper presents a supervisory control scheme based on hybrid systems theory and fuzzy events detection. The fuzzy event detector is a linguistic model, which synthesizes complex relations between process variables and process events incorporating experts' knowledge about the process operation. This kind of detection allows the anticipation of appropriate control actions, which depend upon the selected membership functions used to characterize the process under scrutiny. The proposed supervisory control scheme was successfully implemented for an oxichlorination reactor in a vinyl monomer plant. This implementation has allowed improvement of reactor stability and reduction of raw material consumption.

  18. Design study of lead bismuth cooled fast reactors and capability of natural circulation

    NASA Astrophysics Data System (ADS)

    Oktamuliani, Sri; Su'ud, Zaki

    2015-09-01

    A preliminary study designs SPINNOR (Small Power Reactor, Indonesia, No On-Site Refueling) liquid metal Pb-Bi cooled fast reactors, fuel (U, Pu)N, 150 MWth have been performed. Neutronic calculation uses SRAC which is designed cylindrical core 2D (R-Z) 90 × 135 cm, on the core fuel composed of heterogeneous with percentage difference of PuN 10, 12, 13% and the result of calculation is effective neutron multiplication 1.0488. Power density distribution of the output SRAC is generated for thermal hydraulic calculation using Delphi based on Pascal language that have been developed. The research designed a reactor that is capable of natural circulation at inlet temperature 300 °C with variation of total mass flow rate. Total mass flow rate affect pressure drop and temperature outlet of the reactor core. The greater the total mass flow rate, the smaller the outlet temperature, but increase the pressure drop so that the chimney needed more higher to achieve natural circulation or condition of the system does not require a pump. Optimization of the total mass flow rate produces optimal reactor design on the total mass flow rate of 5000 kg/s with outlet temperature 524,843 °C but require a chimney of 6,69 meters.

  19. Design study of lead bismuth cooled fast reactors and capability of natural circulation

    SciTech Connect

    Oktamuliani, Sri Su’ud, Zaki

    2015-09-30

    A preliminary study designs SPINNOR (Small Power Reactor, Indonesia, No On-Site Refueling) liquid metal Pb-Bi cooled fast reactors, fuel (U, Pu)N, 150 MWth have been performed. Neutronic calculation uses SRAC which is designed cylindrical core 2D (R-Z) 90 × 135 cm, on the core fuel composed of heterogeneous with percentage difference of PuN 10, 12, 13% and the result of calculation is effective neutron multiplication 1.0488. Power density distribution of the output SRAC is generated for thermal hydraulic calculation using Delphi based on Pascal language that have been developed. The research designed a reactor that is capable of natural circulation at inlet temperature 300 °C with variation of total mass flow rate. Total mass flow rate affect pressure drop and temperature outlet of the reactor core. The greater the total mass flow rate, the smaller the outlet temperature, but increase the pressure drop so that the chimney needed more higher to achieve natural circulation or condition of the system does not require a pump. Optimization of the total mass flow rate produces optimal reactor design on the total mass flow rate of 5000 kg/s with outlet temperature 524,843 °C but require a chimney of 6,69 meters.

  20. Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

    SciTech Connect

    Beyer, Brian David; Beddingfield, David H; Durst, Philip; Bean, Robert

    2010-01-01

    The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguards criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.

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

    NASA Astrophysics Data System (ADS)

    Schriener, Timothy M.

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

  2. Demonstration of the Anaerobic Fluidized Bed Reactor for Pinkwater Treatment at McAlester Army Ammunition Plant

    DTIC Science & Technology

    2005-03-01

    column with an overall height of 22 ft (6.7 m) and a bed of GAC occupying approximately 11 ft (3.4 m) when expanded. Water was recirculated through the...Approved for public release; distribution is unlimited. ER D C /C ER L TR -0 5- 8 Demonstration of the Anaerobic Fluidized Bed Reactor for...Anaerobic Fluidized Bed Reactor for Pinkwater Treatment at McAlester Army Ammunition Plant Stephen W. Maloney and Robert L. Heine Construction

  3. Evaluation of the ECAS open cycle MHD power plant design

    NASA Technical Reports Server (NTRS)

    Seikel, G. R.; Staiger, P. J.; Pian, C. C. P.

    1978-01-01

    The Energy Conversion Alternatives Study (ECAS) MHD/steam power plant is described. The NASA critical evaluation of the design is summarized. Performance of the MHD plant is compared to that of the other type ECAS plant designs on the basis of efficiency and the 30-year levelized cost of electricity. Techniques to improve the plant design and the potential performance of lower technology plants requiring shorter development time and lower development cost are then discussed.

  4. Solar coal gasification - Plant design and economics

    NASA Astrophysics Data System (ADS)

    Aiman, W. R.; Thorsness, C. B.; Gregg, D. W.

    A plant design and economic analysis is presented for solar coal gasification (SCG). Coal pyrolysis and char gasification to form the gasified product are reviewed, noting that the endothermic gasification reactions occur only at temperatures exceeding 1000 K, an energy input of 101-136 kJ/mol of char reformed. Use of solar heat offers the possibility of replacing fuels needed to perform the gasification and the oxygen necessary in order to produce a nitrogen-free product. Reactions, energetics, and byproducts from the gasification of subbituminous coal are modeled for a process analysis code used for the SCG plant. Gas generation is designed to occur in a unit exposed to the solar flux focus from a heliostat field. The SCG gas would have an H2 content of 88%, compared to the 55% offered by the Lurgi process. Initial capital costs for the SCG plant are projected to be 4 times those with the Lurgi process, with equality being achieved when coal costs $4/gJ.

  5. The application of Plant Reliability Data Information System (PRINS) to CANDU reactor

    SciTech Connect

    Hwang, S. W.; Lim, Y. H.; Park, H. C.

    2012-07-01

    As risk-informed applications (RIAs) are actively implanted in the nuclear industry, an issue associated with technical adequacy of Probabilistic Safety Assessment (PSA) arises in its modeling and data sourcing. In Korea, PSA for all Korean NPPs has been completed and KHNP(Korea Hydro and Nuclear Power Plant Company) developed the database called the Plant Reliability Data Information System (PRinS). It has several characteristics that distinguish it from other database system such as NPRDs (INPO,1994), PRIS (IAEA), and SRDF (EdF). This database has the function of systematic data management such as automatic data-gathering, periodic data deposition and updating, statistical analysis including Bayesian method, and trend analysis of failure rate or unavailability. In recent PSA for CANDU reactor, the component failure data of EPRI ALWR URD and Component Reliability Database were preferentially used as generic data set. The error factor for most component failure data was estimated by using the information NUREG/CR-4550 and NUREG/CR-4639. Also, annual trend analysis was performed for the functional losses of components using the statistical analysis and chart module of PRinS. Furthermore, the database has been updated regularly and maintained as a living program to reflect the current status. This paper presents the failure data analysis using PRinS which provides Bayesian analysis on main components in the CANDU reactor. (authors)

  6. Fluid modeling and design of gas channels of solar non-stoichiometric redox reactor

    NASA Astrophysics Data System (ADS)

    Kedlaya, Aditya

    The present numerical study in FLUENT analyzes the fluid flow field within a solar powered reactor designed for syngas production. The thermochemical reactor is based on continuous cycling of cerium oxide (ceria) in a non-stoichiometric reduction/oxidation cycle. The reactor uses a hollow cylinder of porous ceria which rotates through a high-temperature zone, by exposure to concentrated sunlight and partially reduced in an inert atmosphere due to flow of the sweep gas (N2), and then through a lower temperature zone where the reduced ceria is re-oxidized with a flow of CO2 and/or H2O, to produce CO and/or H2. In terms of fluid flow modeling, the issue of crossover of species (leakage) within the reactor is critical for proper functioning of the reactor. The first part of the work relates to the geometry and placement of the inlet/outlet gas channels for the reactor optimized to minimize crossover of the species. This is done by conducting a parametric study of geometric variables associated with the inlet/outlet geometry. A simplified 2D fluid flow reactor model which incorporates multi-species flow is used for this study. Further, 2D and 3D reactor models which capture the internal structure more accurately are used to refine the inlet/outlet design. The optimized reactor model is found to have an O2 crossover of 2%-6% and oxidizer crossover of 8%-21% at different flow rates of the sweep gas and the oxidizer studied. In the second part of the work, the reactor model is simulated under varying test conditions. Different working conditions include morphologies of the reactive material, rotational speed of the ceria ring and the recuperator, flow rates of sweep gas and the oxidizer, types of oxidizer (CO2, H2O). The 3D reactor model is also tested using one, two and three discrete inlet/outlet ports and compared with slot configuration.

  7. Lunar in-core thermionic nuclear reactor power system conceptual design

    NASA Technical Reports Server (NTRS)

    Mason, Lee S.; Schmitz, Paul C.; Gallup, Donald R.

    1991-01-01

    This paper presents a conceptual design of a lunar in-core thermionic reactor power system. The concept consists of a thermionic reactor located in a lunar excavation with surface mounted waste heat radiators. The system was integrated with a proposed lunar base concept representative of recent NASA Space Exploration Initiative studies. The reference mission is a permanently-inhabited lunar base requiring a 550 kWe, 7 year life central power station. Performance parameters and assumptions were based on the Thermionic Fuel Element (TFE) Verification Program. Five design cases were analyzed ranging from conservative to advanced. The cases were selected to provide sensitivity effects on the achievement of TFE program goals.

  8. Design and Testing of Vacuum Breaker Check Valve for Simplified Boiling Water Reactor

    SciTech Connect

    Ishii, M.; Xu, Y.; Revankar, S.T.

    2002-07-01

    A new design of the vacuum breaker check valve was developed to replace the mechanical valve in a simplified boiling water reactor. Scaling and design calculations were performed to obtain the geometry of new passive hydraulic vacuum breaker check valve. In order to check the valve performance, a RELAP5 model of the simplified boiling water reactor system with the new valve was developed. The valve was implemented in an integral facility, PUMA and was tested for large break loss of coolant accident. (authors)

  9. Lunar in-core thermionic nuclear reactor power system conceptual design

    SciTech Connect

    Mason, L.S. ); Schmitz, P.C. ); Gallup, D.R. )

    1991-01-05

    This paper presents a conceptual design of a lunar in-core thermionic reactor power system. The concept consists of a thermionic reactor located in a lunar excavation with surface mounted waste heat radiators. The system was integrated with a proposed lunar base concept representative of recent NASA Space Explortion Initiative studies. The reference mission is a permanently-inhabited lunar base requiring a 550 kWe, 7 year life central power station. Performance parameters and assumptions were based on the Thermionic Fuel Element (TFE) Verification Program. Five design cases were analyzed ranging from conservative to advanced. The cases were selected to provide sensitivity effects on the achievement of TFE program goals.

  10. Commercial-Scale Performance Predictions for High-Temperature Electrolysis Plants Coupled to Three Advanced Reactor Types

    SciTech Connect

    M. G. McKellar; J. E. O'Brien; J. S. Herring

    2007-09-01

    This report 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 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.

  11. The design of solar tower power plants

    NASA Astrophysics Data System (ADS)

    Gretz, J.

    The conversion of solar energy into electricity in solar thermal tower power plants is examined. Mirrors attached to mobile, sun-following heliostats concentrate solar rays into the opening of a receiver mounted on a tower. In the receiver, the radiant energy is absorbed by a system of pipes filled with a flowing material which is heated and drives a turbogenerator directly or via a heat exchanger. It is shown that the optics involved in this concept preclude the optimization of the pipe material, since the local distribution of rays in the heater of tower power plants varies diurnally and annually. This requires each pipe section to be designed for maximum stress, even though that stress occurs only at brief intervals during the day.

  12. Comparison of CRBR design-basis events with those of foreign LMFBR plants

    SciTech Connect

    Agrawal, A.K.

    1983-04-01

    As part of the Construction Permit (CP) review of the Clinch River Breeder Reactor Plant (CRBR), the Brookhaven National Laboratory was asked to compare the Design Basis Accidents that are considered in CRBR Preliminary Safety Analysis Report with those of the foreign contemporary plants (PHENIX, SUPER-PHENIX, SNR-300, PFR, and MONJU). A brief introductory review of any special or unusual characteristics of these plants is given. This is followed by discussions of the design basis accidents and their acceptance criteria. In spite of some discrepancies due either to semantics or to licensing decisions, there appears to be a considerable degree of unanimity in the selection (definition) of DBAs in all of these plants.

  13. An autonomous long-term fast reactor system and the principal design limitations of the concept

    NASA Astrophysics Data System (ADS)

    Tsvetkova, Galina Valeryevna

    The objectives of this dissertation were to find a principal domain of promising and technologically feasible reactor physics characteristics for a multi-purpose, modular-sized, lead-cooled, fast neutron spectrum reactor fueled with an advanced uranium-transuranic-nitride fuel and to determine the principal limitations for the design of an autonomous long-term multi-purpose fast reactor (ALM-FR) within the principal reactor physics characteristic domain. The objectives were accomplished by producing a conceptual design for an ALM-FR and by analysis of the potential ALM-FR performance characteristics. The ALM-FR design developed in this dissertation is based on the concept of a secure transportable autonomous reactor for hydrogen production (STAR-H2) and represents further refinement of the STAR-H2 concept towards an economical, proliferation-resistant, sustainable, multi-purpose nuclear energy system. The development of the ALM-FR design has been performed considering this reactor within the frame of the concept of a self-consistent nuclear energy system (SCNES) that satisfies virtually all of the requirements for future nuclear energy systems: efficient energy production, safety, self-feeding, non-proliferation, and radionuclide burning. The analysis takes into consideration a wide range of reactor design aspects including selection of technologically feasible fuels and structural materials, core configuration optimization, dynamics and safety of long-term operation on one fuel loading, and nuclear material non-proliferation. Plutonium and higher actinides are considered as essential components of an advanced fuel that maintains long-term operation. Flexibility of the ALM-FR with respect to fuel compositions is demonstrated acknowledging the principal limitations of the long-term burning of plutonium and higher actinides. To ensure consistency and accuracy, the modeling has been performed using state-of-the-art computer codes developed at Argonne National

  14. Design Options to Reduce Development Cost of First Generation Surface Reactors

    SciTech Connect

    Poston, David I.; Marcille, Thomas F.

    2006-01-20

    Low-power surface reactors have the potential to have the lowest development cost of any space reactor application, primarily because system alpha (mass/kg) is not of utmost importance and mission lifetimes do not have to be a decade or more. Even then, the development cost of a surface reactor can vary substantially depending on the performance requirements (e.g. mass, power, lifetime, reliability) and technical development risk deemed acceptable by the end-user. It is important for potential users to be aware of these relationships before they determine their future architecture (i.e. decide what they need). Generally, the greatest potential costs of a space reactor program are a nuclear-powered ground test and extensive material development campaigns, so it is important to consider options that can minimize the need for or complexity of such tasks. The intended goal of this paper is to inform potential surface reactor users of the potential sensitivities of surface reactor development cost to design requirements, and areas where technical risk can be traded with development cost.

  15. Effect of hydraulic retention time on inorganic nutrient recovery and biodegradable organics removal in a biofilm reactor treating plant biomass leachate

    NASA Technical Reports Server (NTRS)

    Krumins, Valdis; Hummerick, Mary; Levine, Lanfang; Strayer, Richard; Adams, Jennifer L.; Bauer, Jan

    2002-01-01

    A fixed-film (biofilm) reactor was designed and its performance was determined at various retention times. The goal was to find the optimal retention time for recycling plant nutrients in an advanced life support system, to minimize the size, mass, and volume (hold-up) of a production model. The prototype reactor was tested with aqueous leachate from wheat crop residue at 24, 12, 6, and 3 h hydraulic retention times (HRTs). Biochemical oxygen demand (BOD), nitrates and other plant nutrients, carbohydrates, total phenolics, and microbial counts were monitored to characterize reactor performance. BOD removal decreased significantly from 92% at the 24 h HRT to 73% at 3 h. Removal of phenolics was 62% at the 24 h retention time, but 37% at 3 h. Dissolved oxygen concentrations, nitric acid consumption, and calcium and magnesium removals were also affected by HRT. Carbohydrate removals, carbon dioxide (CO2) productions, denitrification, potassium concentrations, and microbial counts were not affected by different retention times. A 6 h HRT will be used in future studies to determine the suitability of the bioreactor effluent for hydroponic plant production.

  16. The Conceptual Design of an Integrated Nuclearhydrogen Production Plant Using the Sulfur Cycle Water Decomposition System

    NASA Technical Reports Server (NTRS)

    Farbman, G. H.

    1976-01-01

    A hydrogen production plant was designed based on a hybrid electrolytic-thermochemical process for decomposing water. The sulfur cycle water decomposition system is driven by a very high temperature nuclear reactor that provides 1,283 K helium working gas. The plant is sized to approximately ten million standard cubic meters per day of electrolytically pure hydrogen and has an overall thermal efficiently of 45.2 percent. The economics of the plant were evaluated using ground rules which include a 1974 cost basis without escalation, financing structure and other economic factors. Taking into account capital, operation, maintenance and nuclear fuel cycle costs, the cost of product hydrogen was calculated at $5.96/std cu m for utility financing. These values are significantly lower than hydrogen costs from conventional water electrolysis plants and competitive with hydrogen from coal gasification plants.

  17. Modeling of the Reactor Core Isolation Cooling Response to Beyond Design Basis Operations - Interim Report

    SciTech Connect

    Ross, Kyle; Cardoni, Jeffrey N.; Wilson, Chisom Shawn; Morrow, Charles; Osborn, Douglas; Gauntt, Randall O.

    2015-12-01

    Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine

  18. Mixed enrichment core design for the NC State University PULSTAR Reactor

    SciTech Connect

    Mayo, C.W.; Verghese, K.; Huo, Y.G.

    1997-12-01

    The North Carolina State University PULSTAR Reactor license was renewed for an additional 20 years of operation on April 30, 1997. The relicensing period added additional years to the facility operating time through the end of the second license period, increasing the excess reactivity needs as projected in 1988. In 1995, the Nuclear Reactor Program developed a strategic plan that addressed the future maintenance, development, and utilization of the facility. Goals resulting from this plan included increased academic utilization of the facility in accordance with its role as a university research facility, and increased industrial service use in accordance with the mission of a land grant university. The strategic plan was accepted, and it is the intent of the College of Engineering to operate the PULSTAR Reactor as a going concern through at least the end of the current license period. In order to reach the next relicensing review without prejudice due to low excess reactivity, it is desired to maintain sufficient excess reactivity so that, if relicensed again, the facility could continue to operate without affecting users until new fuel assistance was provided. During the NC State University license renewal, the operation of the PULSTAR Reactor at the State University of New York at Buffalo (SUNY Buffalo) was terminated. At that time, the SUNY Buffalo facility had about 240 unused PULSTAR Reactor fuel pins with 6% enrichment. The objective of the work reported here was to develop a mixed enrichment core design for the NC State University PULSTAR reactor which would: (1) demonstrate that 6% enriched SUNY buffalo fuel could be used in the NC State University PULSTAR Reactor within the existing technical specification safety limits for core physics parameters; (2) show that use of this fuel could permit operating the NC State University PULSTAR Reactor to 2017 with increased utilization; and (3) assure that the decision whether or not to relicense the facility would

  19. Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor

    SciTech Connect

    Daniel M. Wachs; Richard G. Ambrosek; Gray Chang; Mitchell K. Meyer

    2006-10-01

    Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progress toward element testing will be reviewed.

  20. Photolytic treatment of atrazine-contaminated water: products, kinetics, and reactor design.

    PubMed

    Ye, Xuejun; Chen, Daniel; Li, Kuyen; Wang, Bin; Hopper, Jack

    2007-08-01

    This study investigates the products, kinetics, and reactor design of atrazine photolysis under 254-nm ultraviolet-C (UVC) irradiation. With an initial atrazine concentration of 60 microg/L (60 ppbm), only two products remain in detectable levels. Up to 77% of decomposed atrazine becomes hydroxyatrazine, the major product. Both atrazine and hydroxyatrazine photodecompose following the first-order rate equation, but the hydroxyatrazine photodecomposition rate is significantly slower than that of atrazine. For atrazine photodecomposition, the rate constant is proportional to the square of UVC output, but inversely proportional to the reactor volume. For a photochemical reactor design, a series of equations are proposed to calculate the needed UVC output power, water treatment capacity, and atrazine outlet concentration.

  1. Reactor Pressure Vessel Temperature Analysis for Prismatic and Pebble-Bed VHTR Designs

    SciTech Connect

    H. D. Gougar; C. B. Davis

    2006-04-01

    Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code.

  2. Conceptual Design of a Very High Temperature Pebble-Bed Reactor

    SciTech Connect

    Hans D. Gougar; A. M. Ougouag; Richard M. Moore; W. K. Terry

    2003-11-01

    Efficient electricity and hydrogen production distinguish the Very High Temperature Reactor as the leading Generation IV advanced concept. This graphite-moderated, helium-cooled reactor achieves a requisite high outlet temperature while retaining the passive safety and proliferation resistance required of Generation IV designs. Furthermore, a recirculating pebble-bed VHTR can operate with minimal excess reactivity to yield improved fuel economy and superior resistance to ingress events. Using the PEBBED code developed at the INEEL, conceptual designs of 300 megawatt and 600 megawatt (thermal) Very High Temperature Pebble-Bed Reactors have been developed. The fuel requirements of these compare favorably to the South African PBMR. Passive safety is confirmed with the MELCOR accident analysis code.

  3. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    SciTech Connect

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D.

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  4. Design Concept for a Nuclear Reactor-Powered Mars Rover

    NASA Technical Reports Server (NTRS)

    Elliott, John; Poston, Dave; Lipinski, Ron

    2007-01-01

    A report presents a design concept for an instrumented robotic vehicle (rover) to be used on a future mission of exploration of the planet Mars. The design incorporates a nuclear fission power system to provide long range, long life, and high power capabilities unachievable through the use of alternative solar or radioisotope power systems. The concept described in the report draws on previous rover designs developed for the 2009 Mars Science laboratory (MSL) mission to minimize the need for new technology developments.

  5. Neutronic design studies for an unattended, low power reactor

    SciTech Connect

    Palmer, R.G.; Durkee, J.W. Jr.

    1986-01-01

    The Los Alamos National Laboratory is involved in the design and demonstrations of a small, long-lived nuclear heat and electric power source for potential applications at remote sites where alternate fossil energy systems would not be cost effective. This paper describes the neutronic design analysis that was performed to arrive at two conceptual designs, one using thermoelectric conversion, the other using an organic Rankine cycle. To meet the design objectives and constraints a number of scoping and optimization studies were carried out. The results of calculations of control worths, temperature coefficients of reactivity and fuel depletion effects are reported.

  6. Validation of FSP Reactor Design with Sensitivity Studies of Beryllium-Reflected Critical Assemblies

    SciTech Connect

    John D. Bess; Margaret A. Marshall

    2013-02-01

    The baseline design for space nuclear power is a fission surface power (FSP) system: sodium-potassium (NaK) cooled, fast spectrum reactor with highly-enriched-uranium (HEU)-O2 fuel, stainless steel (SS) cladding, and beryllium reflectors with B4C control drums. Previous studies were performed to evaluate modeling capabilities and quantify uncertainties and biases associated with analysis methods and nuclear data. Comparison of Zero Power Plutonium Reactor (ZPPR)-20 benchmark experiments with the FSP design indicated that further reduction of the total design model uncertainty requires the reduction in uncertainties pertaining to beryllium and uranium cross-section data. Further comparison with three beryllium-reflected HEU-metal benchmark experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) concluded the requirement that experimental validation data have similar cross section sensitivities to those found in the FSP design. A series of critical experiments was performed at ORCEF in the 1960s to support the Medium Power Reactor Experiment (MPRE) space reactor design. The small, compact critical assembly (SCCA) experiments were graphite- or beryllium-reflected assemblies of SS-clad, HEU-O2 fuel on a vertical lift machine. All five configurations were evaluated as benchmarks. Two of the five configurations were beryllium reflected, and further evaluated using the sensitivity and uncertainty analysis capabilities of SCALE 6.1. Validation of the example FSP design model was successful in reducing the primary uncertainty constituent, the Be(n,n) reaction, from 0.28 %dk/k to 0.0004 %dk/k. Further assessment of additional reactor physics measurements performed on the SCCA experiments may serve to further validate FSP design and operation.

  7. Thermal Response of the Hybrid Loop-Pool Design for Sodium Cooled Faster Reactors

    SciTech Connect

    Zhang, Hongbin; Zhao, Haihua; Davis, Cliff

    2008-09-01

    An innovative hybrid loop-pool design for the sodium cooled fast reactor (SFR) has been recently proposed with the primary objective of achieving cost reduction and safety enhancement. With the hybrid loop-pool design, closed primary loops are immersed in a secondary buffer tank. This design takes advantage of features from conventional both pool and loop designs to further improve economics and safety. This paper will briefly introduce the hybrid loop-pool design concept and present the calculated thermal responses for unproctected (without reactor scram) loss of forced circulation (ULOF) transients using RELAP5-3D. The analyses examine both the inherent reactivity shutdown capability and decay heat removal performance by passive safety systems.

  8. Summary of the Advanced Reactor Design Criteria (ARDC) Phase 2 Activities

    SciTech Connect

    Holbrook, Mark Raymond

    2015-09-01

    This report provides an end-of-year summary reflecting the progress and status of proposed regulatory design criteria for advanced non-LWR designs in accordance with the Level 3 milestone in M3AT-15IN2001017 in work package AT-15IN200101. These criteria have been designated as ARDC, and they provide guidance to future applicants for addressing the GDC that are currently applied specifically to LWR designs. The report provides a summary of Phase 2 activities related to the various tasks associated with ARDC development and the subsequent development of example adaptations of ARDC for Sodium Fast Reactor (SFR) and modular High Temperature Gas-cooled Reactor (HTGR) designs.

  9. American National Standard: design basis for protection of light water nuclear power plants against effects of postulated pipe rupture

    SciTech Connect

    Not Available

    1980-12-31

    This standard addresses the design bases for light water reactor, nuclear power plant structures and components essential for the protection of public health and safety from the potential adverse effects of pipe whip, jet impingement, pressurization of compartments outside containment, environmental conditions and flooding associated with a postulated pipe rupture. The design bases for missile protection and the design bases for containment pressurization are not within this standard.

  10. Design improvement and performance evaluation of solar photocatalytic reactor for industrial effluent treatment.

    PubMed

    Nair, Ranjith G; Bharadwaj, P J; Samdarshi, S K

    2016-12-01

    This work reports the details of the design components and materials used in a linear compound parabolic trough reactor constructed with an aim to use the photocatalyst for solar photocatalytic applications. A compound parabolic trough reactor has been designed and engineered to exploit both UV and visible part of the solar irradiation. The developed compound parabolic trough reactor could receive almost 88% of UV radiation along with a major part of visible radiation. The performance of the reactor has been evaluated in terms of degradation of a probe pollutant using the parameters such as rate constant, residence time and photonic efficiency. An attempt has been made to assess the performance in different ranges of solar spectrum. Finally the developed reactor has been employed for the photocatalytic treatment of a paper mill effluent using Degussa P25 as the photocatalyst. The paper mill effluent collected from Nagaon paper mill, Assam, India has been treated under both batch mode and continuous mode using Degussa P25 photocatalyst under artificial and natural solar radiation, respectively. The photocatalytic degradation kinetics of the paper mill effluent has been determined using the reduction in total organic carbon (TOC) values of the effluent.

  11. Nuclear Island Engineering MHTGR [Modular High-Temperature Gas-cooled Reactor] preliminary and final designs. Technical progress report, December 12, 1988--September 30, 1989

    SciTech Connect

    1989-12-01

    This report summarizes the Department of Energy (DOE)-funded work performed by General Atomics (GA) under the Nuclear Island Engineering (NIE)-Modular High-Temperature Gas-cooled Reactor (MHTGR) Preliminary and Final Designs Contract DE-AC03-89SF17885 for the period December 12, 1988 through September 30, 1989. This reporting period is the first (partial) fiscal year of the 5-year contract performance period. The objective of DOE`s MHTGR program is to advance the design from the conceptual design phase into preliminary design and then on to final design in support of the Nuclear Regulatory Commission`s (NRC`s) design review and approval of the MHTGR Design Team, is focused on the Nuclear Island portion of the technology and design, primarily in the areas of the reactor and internals, fuel characteristics and fuel fabrication, helium services systems, reactor protection, shutdown cooling, circulator design, and refueling system. Maintenance and implementation of the functional methodology, plant-level analysis, support for probabilistic risk assessment, quality assurance, operations, and reliability/availability assessments are included in GA`s scope of work.

  12. DEVELOPMENT AND TESTING OF FAULT-DIAGNOSIS ALGORITHMS FOR REACTOR PLANT SYSTEMS

    SciTech Connect

    Grelle, Austin L.; Park, Young S.; Vilim, Richard B.

    2016-06-26

    Argonne National Laboratory is further developing fault diagnosis algorithms for use by the operator of a nuclear plant to aid in improved monitoring of overall plant condition and performance. The objective is better management of plant upsets through more timely, informed decisions on control actions with the ultimate goal of improved plant safety, production, and cost management. Integration of these algorithms with visual aids for operators is taking place through a collaboration under the concept of an operator advisory system. This is a software entity whose purpose is to manage and distill the enormous amount of information an operator must process to understand the plant state, particularly in off-normal situations, and how the state trajectory will unfold in time. The fault diagnosis algorithms were exhaustively tested using computer simulations of twenty different faults introduced into the chemical and volume control system (CVCS) of a pressurized water reactor (PWR). The algorithms are unique in that each new application to a facility requires providing only the piping and instrumentation diagram (PID) and no other plant-specific information; a subject-matter expert is not needed to install and maintain each instance of an application. The testing approach followed accepted procedures for verifying and validating software. It was shown that the code satisfies its functional requirement which is to accept sensor information, identify process variable trends based on this sensor information, and then to return an accurate diagnosis based on chains of rules related to these trends. The validation and verification exercise made use of GPASS, a one-dimensional systems code, for simulating CVCS operation. Plant components were failed and the code generated the resulting plant response. Parametric studies with respect to the severity of the fault, the richness of the plant sensor set, and the accuracy of sensors were performed as part of the validation

  13. Innovative Offshore Wind Plant Design Study

    SciTech Connect

    Hurley, William L.; Nordstrom, Charles J.; Morrison, Brent J.

    2013-12-18

    Technological advancements in the Glosten PelaStar floating wind turbine system have led to projected cost of energy (COE) reductions from today’s best-in-class offshore wind systems. The PelaStar system is projected to deliver a COE that is 35% lower than that delivered by the current offshore wind plants. Several technology developments have been achieved that directly target significant cost of energy reductions. These include: Application of state-of-the-art steel construction materials and methods, including fatigue-resistant welding techniques and technologies, to reduce hull steel weight; Advancements in synthetic fiber tendon design for the mooring system, which are made possible by laboratory analysis of full-scale sub-rope specimens; Investigations into selected anchor technologies to improve anchor installation methods; Refinement of the installation method, specifically through development of the PelaStar Support Barge design. Together, these technology developments drive down the capital cost and operating cost of offshore wind plants and enable access to superb wind resources in deep water locations. These technology developments also reduce the uncertainty of the PelaStar system costs, which increases confidence in the projected COE reductions.

  14. Recommended practices in elevated temperature design: A compendium of breeder reactor experiences (1970-1986): An overview

    SciTech Connect

    Wei, B.C.; Cooper, W.L. Jr.; Dhalla, A.K.

    1987-09-01

    Significant experiences have been accumulated in the establishment of design methods and criteria applicable to the design of Liquid Metal Fast Breeder Reactor (LMFBR) components. The Subcommittee of the Elevated Temperature Design under the Pressure Vessel Research Council (PVRC) has undertaken to collect, on an international basis, design experience gained, and the lessons learned, to provide guidelines for next generation advanced reactor designs. This paper shall present an overview and describe the highlights of the work.

  15. Induced Radioactivity and Waste Classification of Reactor Zone Components of the Chernobyl Nuclear Power Plant Unit 1 After Final Shutdown

    SciTech Connect

    Bylkin, Boris K.; Davydova, Galina B.; Zverkov, Yuri A.; Krayushkin, Alexander V.; Neretin, Yuri A.; Nosovsky, Anatoly V.; Seyda, Valery A.; Short, Steven M.

    2001-10-15

    The dismantlement of the reactor core materials and surrounding structural components is a major technical concern for those planning closure and decontamination and decommissioning of the Chernobyl Nuclear Power Plant (NPP). Specific issues include when and how dismantlement should be accomplished and what the radwaste classification of the dismantled system would be at the time it is disassembled. Whereas radiation levels and residual radiological characteristics of the majority of the plant systems are directly measured using standard radiation survey and radiochemical analysis techniques, actual measurements of reactor zone materials are not practical due to high radiation levels and inaccessibility. For these reasons, neutron transport analysis was used to estimate induced radioactivity and radiation levels in the Chernobyl NPP Unit 1 reactor core materials and structures.Analysis results suggest that the optimum period of safe storage is 90 to 100 yr for the Unit 1 reactor. For all of the reactor components except the fuel channel pipes (or pressure tubes), this will provide sufficient decay time to allow unlimited worker access during dismantlement, minimize the need for expensive remote dismantlement, and allow for the dismantled reactor components to be classified as low- or medium-level radioactive waste. The fuel channel pipes will remain classified as high-activity waste requiring remote dismantlement for hundreds of years due to the high concentration of induced {sup 63}Ni in the Zircaloy pipes.

  16. Spring design for use in the core of a nuclear reactor

    DOEpatents

    Willard, Jr., H. James

    1993-01-01

    A spring design particularly suitable for use in the core of a nuclear reactor includes one surface having a first material oriented in a longitudinal direction, and another surface having a second material oriented in a transverse direction. The respective surfaces exhibit different amounts of irraditation induced strain.

  17. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    SciTech Connect

    Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

    2004-10-06

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are

  18. RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

    SciTech Connect

    Hongbin Zhang; Haihua Zhao; Cliff Davis

    2008-06-01

    An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heat from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet

  19. Design Strategies for Optically-Accessible, High-Temperature, High-Pressure Reactor

    SciTech Connect

    S. F. Rice; R. R. Steeper; C. A. LaJeunesse; R. G. Hanush; J. D. Aiken

    2000-02-01

    The authors have developed two optical cell designs for high-pressure and high-temperature fluid research: one for flow systems, and the other for larger batch systems. The flow system design uses spring washers to balance the unequal thermal expansions of the reactor and the window materials. A typical design calculation is presented showing the relationship between system pressure, operating temperature, and torque applied to the window-retaining nut. The second design employs a different strategy more appropriate for larger windows. This design uses two seals: one for the window that benefits from system pressure, and a second one that relies on knife-edge, metal-to-metal contact.

  20. Design strategies for optically-accessible, high-temperature, high-pressure reactor

    SciTech Connect

    S. F. Rice; R. R. Steeper; C. A. LaJeunesse; R. G. Hanush; J. D. Aiken

    2000-02-01

    The authors have developed two optical cell designs for high-pressure and high-temperature fluid research: one for flow systems, and the other for larger batch systems. The flow system design uses spring washers to balance the unequal thermal expansions of the reactor and the window materials. A typical design calculation is presented showing the relationship between system pressure, operating temperature, and torque applied to the window-retaining nut. The second design employs a different strategy more appropriate for larger windows. This design uses two seals: one for the window that benefits from system pressure, and a second one that relies on knife-edge, metal-to-metal contact.

  1. Thermal-hydraulics design comparisons for the tandem mirror hybrid reactor blanket

    SciTech Connect

    Wong, C.P.C.; Yang, Y.S.; Schultz, K.R.

    1980-09-01

    The Tandem Mirror Hybrid Reactor (TMHR) is a cylindrical reactor, and the fertile materials and tritium breeding fuel elements can be arranged with radial or axial orientation in the blanket module. Thermal-hydraulics performance comparisons were made between plate, axial rod and radial rod fuel geometrices. The three configurations result in different coolant/void fractions and different clad/structure fractions. The higher void fraction in the two rod designs means that these blankets will have to be thicker than the plate design blanket in order to achieve the same level of nuclear interactions. Their higher structural fractions will degrade the uranium breeding ratio and energy multiplication factor of the design. One difficulty in the thermal-hydraulics analysis of the plate design was caused by the varying energy multiplication of the blanket during the lifetime of the plate which forced the use of designs that operated in the transition flow regime at some point during life. To account for this, an approach was adopted from Gas Cooled Fast Reactor (GCFR) experience for the pressure drop calculation and the corresponding heat transfer coefficient that was used for the film drop thermal calculation. Because of the superior nuclear performance, the acceptable thermal-hydraulic characteristics and the mechanical design feasibility, the plate geometry concept was chosen for the reference gas-cooled TMHR blanket design.

  2. Conceptual design of a pressure tube light water reactor with variable moderator control

    SciTech Connect

    Rachamin, R.; Fridman, E.; Galperin, A.

    2012-07-01

    This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

  3. Fuel performance models for high-temperature gas-cooled reactor core design

    SciTech Connect

    Stansfield, O.M.; Simon, W.A.; Baxter, A.M.

    1983-09-01

    Mechanistic fuel performance models are used in high-temperature gas-cooled reactor core design and licensing to predict failure and fission product release. Fuel particles manufactured with defective or missing SiC, IPyC, or fuel dispersion in the buffer fail at a level of less than 5 x 10/sup -4/ fraction. These failed particles primarily release metallic fission products because the OPyC remains intact on 90% of the particles and retains gaseous isotopes. The predicted failure of particles using performance models appears to be conservative relative to operating reactor experience.

  4. Contribution to the validation of MCNP neutronics design of the advanced neutron source reactor

    SciTech Connect

    Rubio, G.A.; Ougouag, A.M.; Wemple, C.A.; Ryskamp, J.M. )

    1993-01-01

    In the research and development plan of the advanced neutron source (ANS) reactor, one of the steps planned for neutronics design methods validation is to compare computational model results and experimental results for a critical facility exhibiting similarities with the ANS reactor. One such facility is the FOEHN experiment. In this paper, an MCNP model of the FOEHN experiment is developed, and its results are compared with experimental data from the literature. The MCNP models reproduces measured quantities of interest with a high level of agreement.

  5. Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration. CEDT Phase 1 Preliminary Design Documentation

    SciTech Connect

    Sanchez, Rene Gerardo; Hutchinson, Jesson D.; Mcclure, Patrick Ray; Myers, William L.

    2015-08-20

    The intent of the integral experiment request IER 299 (called KiloPower by NASA) is to assemble and evaluate the operational performance of a compact reactor configuration that closely resembles the flight unit to be used by NASA to execute a deep space exploration mission. The reactor design will include heat pipes coupled to Stirling engines to demonstrate how one can generate electricity when extracting energy from a “nuclear generated” heat source. This series of experiments is a larger scale follow up to the DUFF series of experiments1,2 that were performed using the Flat-Top assembly.

  6. Advanced Test Reactor In-Canal Ultrasonic Scanner: Experiment Design and Initial Results on Irradiated Plates

    SciTech Connect

    D. M. Wachs; J. M. Wight; D. T. Clark; J. M. Williams; S. C. Taylor; D. J. Utterbeck; G. L. Hawkes; G. S. Chang; R. G. Ambrosek; N. C. Craft

    2008-09-01

    An irradiation test device has been developed to support testing of prototypic scale plate type fuels in the Advanced Test Reactor. The experiment hardware and operating conditions were optimized to provide the irradiation conditions necessary to conduct performance and qualification tests on research reactor type fuels for the RERTR program. The device was designed to allow disassembly and reassembly in the ATR spent fuel canal so that interim inspections could be performed on the fuel plates. An ultrasonic scanner was developed to perform dimensional and transmission inspections during these interim investigations. Example results from the AFIP-2 experiment are presented.

  7. Design of a fault diagnosis system for next generation nuclear power plants

    SciTech Connect

    Zhao, K.; Upadhyaya, B.R.; Wood, R.T.

    2004-07-01

    A new design approach for fault diagnosis is developed for next generation nuclear power plants. In the nuclear reactor design phase, data reconciliation is used as an efficient tool to determine the measurement requirements to achieve the specified goal of fault diagnosis. In the reactor operation phase, the plant measurements are collected to estimate uncertain model parameters so that a high fidelity model can be obtained for fault diagnosis. The proposed algorithm of fault detection and isolation is able to combine the strength of first principle model based fault diagnosis and the historical data based fault diagnosis. Principal component analysis on the reconciled data is used to develop a statistical model for fault detection. The updating of the principal component model based on the most recent reconciled data is a locally linearized model around the current plant measurements, so that it is applicable to any generic nonlinear systems. The sensor fault diagnosis and process fault diagnosis are decoupled through considering the process fault diagnosis as a parameter estimation problem. The developed approach has been applied to the IRIS helical coil steam generator system to monitor the operational performance of individual steam generators. This approach is general enough to design fault diagnosis systems for the next generation nuclear power plants. (authors)

  8. Fundamentals of Electrical Propulsion Plant Design,

    DTIC Science & Technology

    1982-04-06

    Classification. Turboelectric Propulsion Plants ( TEGU ). Diesel- Electric Propulsion Plants (DEGU). § 9.3 Excitation and Control Systems with...is evident, for example, from /431 the data in Table 9.1. Table 9.1 Turboelectric Propulsion Plant ( TEGU ) Basic Characteristics Ship Name Plant Output...generators for GEU: nonsalient-pole machinery in turboelectric propulsion plants ( TEGU ) and salient-pole in diesel electric propulsion plants. Primary

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

    NASA Technical Reports Server (NTRS)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

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

  10. High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems.

    PubMed

    Mahadevan, Vijay S; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

    2014-08-06

    An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.

  11. High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

    PubMed Central

    Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

    2014-01-01

    An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework. PMID:24982250

  12. Options Study Documenting the Fast Reactor Fuels Innovative Design Activity

    SciTech Connect

    Jon Carmack; Kemal Pasamehmetoglu

    2010-07-01

    This document provides presentation and general analysis of innovative design concepts submitted to the FCRD Advanced Fuels Campaign by nine national laboratory teams as part of the Innovative Transmutation Fuels Concepts Call for Proposals issued on October 15, 2009 (Appendix A). Twenty one whitepapers were received and evaluated by an independent technical review committee.

  13. Design of a PID Controller for a PCR Micro Reactor

    ERIC Educational Resources Information Center

    Dinca, M. P.; Gheorghe, M.; Galvin, P.

    2009-01-01

    Proportional-integral-derivative (PID) controllers are widely used in process control, and consequently they are described in most of the textbooks on automatic control. However, rather than presenting the overall design process, the examples given in such textbooks are intended to illuminate specific focused aspects of selection, tuning and…

  14. Problems and prospects connected with development of high-temperature filtration technology at nuclear power plants equipped with VVER-1000 reactors

    NASA Astrophysics Data System (ADS)

    Shchelik, S. V.; Pavlov, A. S.

    2013-07-01

    Results of work on restoring the service properties of filtering material used in the high-temperature reactor coolant purification system of a VVER-1000 reactor are presented. A quantitative assessment is given to the effect from subjecting a high-temperature sorbent to backwashing operations carried out with the use of regular capacities available in the design process circuit in the first years of operation of Unit 3 at the Kalinin nuclear power plant. Approaches to optimizing this process are suggested. A conceptual idea about comprehensively solving the problem of achieving more efficient and safe operation of the high-temperature active water treatment system (AWT-1) on a nuclear power industry-wide scale is outlined.

  15. Thermal-Hydraulic Design of a Fluoride High-Temperature Demonstration Reactor

    SciTech Connect

    Carbajo, Juan J; Qualls, A L

    2016-01-01

    INTRODUCTION The Fluoride High-Temperature Reactor (FHR) named the Demonstration Reactor (DR) is a novel reactor concept using molten salt coolant and TRIstructural ISOtropic (TRISO) fuel that is being developed at Oak Ridge National Laboratory (ORNL). The objective of the FHR DR is to advance the technology readiness level of FHRs. The FHR DR will demonstrate technologies needed to close remaining gaps to commercial viability. The FHR DR has a thermal power of 100 MWt, very similar to the SmAHTR, another FHR ORNL concept (Refs. 1 and 2) with a power of 125 MWt. The FHR DR is also a small version of the Advanced High Temperature Reactor (AHTR), with a power of 3400 MWt, cooled by a molten salt and also being developed at ORNL (Ref. 3). The FHR DR combines three existing technologies: (1) high-temperature, low-pressure molten salt coolant, (2) high-temperature coated-particle TRISO fuel, (3) and passive decay heat cooling systems by using Direct Reactor Auxiliary Cooling Systems (DRACS). This paper presents FHR DR thermal-hydraulic design calculations.

  16. Preliminary safety calculations to improve the design of Molten Salt Fast Reactor

    SciTech Connect

    Brovchenko, M.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Capellan, N.; Ghetta, V.; Laureau, A.

    2012-07-01

    Molten salt reactors are liquid fuel reactors so that they are flexible in operation but very different in the safety approach from solid fuel reactors. This study bears on the specific concept named Molten Salt Fast Reactor (MSFR). Since this new nuclear technology is in development, safety is an essential point to be considered all along the R and D studies. This paper presents the first step of the safety approach: the systematic description of the MSFR, limited here to the main systems surrounding the core. This systematic description is the basis on which we will be able to devise accidental scenarios. Thanks to the negative reactivity feedback coefficient, most accidental scenarios lead to reactor shut down. Because of the decay heat generated in the fuel salt, it must be cooled. After the description of the tools developed to calculate the residual heat, the different contributions are discussed in this study. The decay heat of fission products in the MSFR is evaluated to be low (3% of nominal power), mainly due to the reprocessing that transfers the fission products to the gas reprocessing unit. As a result, the contribution of the actinides is significant (0.5% of nominal power). The unprotected loss of heat sink transients are studied in this paper. It appears that slow transients are favorable (> 1 min) to minimize the temperature increase of the fuel salt. This work will be the basis of further safety studies as well as an essential parameter for the design of the draining system. (authors)

  17. Design of Energy Storage Reactors for Dc-To-Dc Converters. Ph.D. Thesis

    NASA Technical Reports Server (NTRS)

    Chen, D. Y.

    1975-01-01

    Two methodical approaches to the design of energy-storage reactors for a group of widely used dc-to-dc converters are presented. One of these approaches is based on a steady-state time-domain analysis of piecewise-linearized circuit models of the converters, while the other approach is based on an analysis of the same circuit models, but from an energy point of view. The design procedure developed from the first approach includes a search through a stored data file of magnetic core characteristics and results in a list of usable reactor designs which meet a particular converter's requirements. Because of the complexity of this procedure, a digital computer usually is used to implement the design algorithm. The second approach, based on a study of the storage and transfer of energy in the magnetic reactors, leads to a straightforward design procedure which can be implemented with hand calculations. An equation to determine the lower-bound volume of workable cores for given converter design specifications is derived. Using this computer lower-bound volume, a comparative evaluation of various converter configurations is presented.

  18. Applying design principles to fusion reactor configurations for propulsion in space

    NASA Technical Reports Server (NTRS)

    Carpenter, Scott A.; Deveny, Marc E.; Schulze, Norman R.

    1993-01-01

    The application of fusion power to space propulsion requires rethinking the engineering-design solution to controlled-fusion energy. Whereas the unit cost of electricity (COE) drives the engineering-design solution for utility-based fusion reactor configurations; initial mass to low earth orbit (IMLEO), specific jet power (kW(thrust)/kg(engine)), and reusability drive the engineering-design solution for successful application of fusion power to space propulsion. We applied three design principles (DP's) to adapt and optimize three candidate-terrestrial-fusion-reactor configurations for propulsion in space. The three design principles are: provide maximum direct access to space for waste radiation, operate components as passive radiators to minimize cooling-system mass, and optimize the plasma fuel, fuel mix, and temperature for best specific jet power. The three candidate terrestrial fusion reactor configurations are: the thermal barrier tandem mirror (TBTM), field reversed mirror (FRM), and levitated dipole field (LDF). The resulting three candidate space fusion propulsion systems have their IMLEO minimized and their specific jet power and reusability maximized. We performed a preliminary rating of these configurations and concluded that the leading engineering-design solution to space fusion propulsion is a modified TBTM that we call the Mirror Fusion Propulsion System (MFPS).

  19. Core Design Characteristics of the Fluoride Salt-Cooled High Temperature Demonstration Reactor

    SciTech Connect

    Brown, Nicholas R; Qualls, A L; Betzler, Benjamin R; Carbajo, Juan J; Greenwood, Michael Scott; Hale, Richard Edward; Harrison, Thomas J; Powers, Jeffrey J; Robb, Kevin R

    2016-01-01

    Fluoride salt-cooled high temperature reactors (FHRs) are a promising reactor technology option with significant knowledge gaps to implementation. One potential approach to address those technology gaps is via a small-scale demonstration reactor with the goal of increasing the technology readiness level (TRL) of the overall system for the longer term. The objective of this paper is to outline a notional concept for such a system, and to address how the proposed concept would advance the TRL of FHR concepts. Development of the proposed FHR Demonstration Reactor (DR) will enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. Important capabilities that will be demonstrated by building and operating the FHR DR include core design methodologies; fabrication and operation of high temperature reactors; salt procurement, handling, maintenance, and ultimate disposal; salt chemistry control to maximize vessel life; tritium management; heat exchanger performance; pump performance; and reactivity control. The FHR DR is considered part of a broader set of FHR technology development and demonstration efforts, some of which are already underway. Nonreactor test efforts (e.g., heated salt loops or loops using simulant fluids) can demonstrate many technologies necessary for commercial deployment of FHRs. The FHR DR, however, fulfills a crucial role in FHR technology development by advancing the technical maturity and readiness level of the system as a whole.

  20. Unique features of space reactors

    SciTech Connect

    Buden, D.

    1990-01-01

    Space reactors are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K. 8 refs., 3 figs., 1 tab.

  1. The Next Generation Nuclear Plant - Insights Gained from the INEEL Point Design Studies

    SciTech Connect

    Philip E. MacDonald; A. M. Baxter; P. D. Bayless; J. M. Bolin; H. D. Gougar; R. L. Moore; A. M. Ougouag; M. B. Richards; R. L. Sant; J. W. Sterbentz; W. K. Terry

    2004-08-01

    This paper provides the results of an assessment of two possible versions of the Next Generation Nuclear Plant (NGNP), a prismatic fuel type helium gas-cooled reactor and a pebble-bed fuel helium gas reactor. Insights gained regarding the strengths and weaknesses of the two designs are also discussed. Both designs will meet the three basic requirements that have been set for the NGNP: a coolant outlet temperature of 1000 C, passive safety, and a total power output consistent with that expected for commercial high-temperature gas-cooled reactors. Two major modifications of the current Gas Turbine- Modular Helium Reactor (GT-MHR) design were needed to obtain a prismatic block design with a 1000 C outlet temperature: reducing the bypass flow and better controlling the inlet coolant flow distribution to the core. The total power that could be obtained for different core heights without exceeding a peak transient fuel temperature of 1600 °C during a high or low-pressure conduction cooldown event was calculated. With a coolant inlet temperature of 490 °C and 10% nominal core bypass flow, it is estimated that the peak power for a 10-block high core is 686 MWt, for a 12-block high core is 786 MWt, and for a 14-block core is about 889 MWt. The core neutronics calculations showed that the NGNP will exhibit strongly negative Doppler and isothermal temperature coefficients of reactivity over the burnup cycle. In the event of rapid loss of the helium gas, there is negligible core reactivity change. However, water or steam ingress into the core coolant channels can produce a relatively large reactivity effect. Two versions of an annular pebble-bed NGNP have also been developed, a 300 and a 600 MWt module. From this work we learned how to design passively safe pebble bed reactors that produce more than 600 MWt. We also found a way to improve both the fuel utilization and safety by modifying the pebble design (by adjusting the fuel zone radius in the pebble to optimize the fuel

  2. Design of a lunar oxygen production plant

    NASA Technical Reports Server (NTRS)

    Radhakrishnan, Ramalingam

    1990-01-01

    To achieve permanent human presence and activity on the moon, oxygen is required for both life support and propulsion. Lunar oxygen production using resources existing on the moon will reduce or eliminate the need to transport liquid oxygen from earth. In addition, the co-products of oxygen production will provide metals, structural ceramics, and other volatile compounds. This will enable development of even greater self-sufficiency as the lunar outpost evolves. Ilmenite is the most abundant metal-oxide mineral in the lunar regolith. A process involving the reaction of ilmenite with hydrogen at 1000 C to produce water, followed by the electrolysis of this water to provide oxygen and recycle the hydrogen has been explored. The objective of this 1990 Summer Faculty Project was to design a lunar oxygen-production plant to provide 5 metric tons of liquid oxygen per year from lunar soil. The results of this study describe the size and mass of the equipment, the power needs, feedstock quantity and the engineering details of the plant.

  3. ORIGEN-ARP Cross-Section Libraries for Magnox, Advanced Gas-Cooled, and VVER Reactor Designs

    SciTech Connect

    Murphy, BD

    2004-03-10

    Cross-section libraries for the ORIGEN-ARP system were extended to include four non-U.S. reactor types: the Magnox reactor, the Advanced Gas-Cooled Reactor, the VVER-440, and the VVER-1000. Typical design and operational parameters for these four reactor types were determined by an examination of a variety of published information sources. Burnup simulation models of the reactors were then developed using the SAS2H sequence from the Oak Ridge National Laboratory SCALE code system. In turn, these models were used to prepare the burnup-dependent cross-section libraries suitable for use with ORIGEN-ARP. The reactor designs together with the development of the SAS2H models are described, and a small number of validation results using spent-fuel assay data are reported.

  4. A Heterogeneous Sodium Fast Reactor Designed to Transmute Minor Actinide Actinide Waste Isotopes into Plutonium Fuel

    SciTech Connect

    Samuel E. Bays

    2011-02-01

    An axial heterogeneous sodium fast reactor design is developed for converting minor actinide waste isotopes into plutonium fuel. The reactor design incorporates zirconium hydride moderating rods in an axial blanket above the active core. The blanket design traps the active core’s axial leakage for the purpose of transmuting Am-241 into Pu-238. This Pu-238 is then co-recycled with the spent driver fuel to make new driver fuel. Because Pu-238 is significantly more fissile than Am-241 in a fast neutron spectrum, the fissile worth of the initial minor actinide material is upgraded by its preconditioning via transmutation in the axial targets. Because, the Am-241 neutron capture worth is significantly stronger in a moderated epithermal spectrum than the fast spectrum, the axial targets serve as a neutron trap which recovers the axial leakage lost by the active core. The sodium fast reactor proposed by this work is designed as an overall transuranic burner. Therefore, a low transuranic conversion ratio is achieved by a degree of core flattening which increases axial leakage. Unlike a traditional “pancake” design, neutron leakage is recovered by the axial target/blanket system. This heterogeneous core design is constrained to have sodium void and Doppler reactivity worth similar to that of an equivalent homogeneous design. Because minor actinides are irradiated only once in the axial target region; elemental partitioning is not required. This fact enables the use of metal targets with electrochemical reprocessing. Therefore, the irradiation environment of both drivers and targets was constrained to ensure applicability of the established experience database for metal alloy sodium fast reactor fuels.

  5. Analysis of Improved Reference Design for a Nuclear-Driven High Temperature Electrolysis Hydrogen Production Plant

    SciTech Connect

    Edwin A. Harvego; James E. O'Brien; Michael G. McKellar

    2010-06-01

    The use of High Temperature Electrolysis (HTE) for the efficient production of hydrogen without the greenhouse gas emissions associated with conventional fossil-fuel hydrogen production techniques has been under investigation at the Idaho National Engineering Laboratory (INL) for the last several years. The activities at the INL have included the development, testing and analysis of large numbers of solid oxide electrolysis cells, and the analyses of potential plant designs for large scale production of hydrogen using an advanced Very-High Temperature Reactor (VHTR) to provide the process heat and electricity to drive the electrolysis process. The results of these system analyses, using the UniSim process analysis software, have shown that the HTE process, when coupled to a VHTR capable of operating at reactor outlet temperatures of 800 °C to 950 °C, has the potential to produce the large quantities of hydrogen needed to meet future energy and transportation needs with hydrogen production efficiencies in excess of 50%. In addition, economic analyses performed on the INL reference plant design, optimized to maximize the hydrogen production rate for a 600 MWt VHTR, have shown that a large nuclear-driven HTE hydrogen production plant can to be economically competitive with conventional hydrogen production processes, particularly when the penalties associated with greenhouse gas emissions are considered. The results of this research led to the selection in 2009 of HTE as the preferred concept in the U.S. Department of Energy (DOE) hydrogen technology down-selection process. However, the down-selection process, along with continued technical assessments at the INL, has resulted in a number of proposed modifications and refinements to improve the original INL reference HTE design. These modifications include changes in plant configuration, operating conditions and individual component designs. This paper describes the resulting new INL reference design and presents

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

    SciTech Connect

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

    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

  7. The effects of spectral shift absorbers on the design and safety of fast spectrum space reactors

    NASA Astrophysics Data System (ADS)

    King, Jeffrey Charles

    Spectral Shift Absorbers (SSAs) are incorporated into space reactors to maintain them sufficiently subcritical when submerged in seawater or wet sand and subsequently flooded, following a launch abort accident. The effect of four SSAs (samarium-149, europium-151, gadolinium-155, and gadolinium-157) on the submersion criticality, operation, and temperature reactivity feedback of the thermal spectrum reactors developed in the Systems for Nuclear Auxilary Power (SNAP) program is extensively documented. Recent work on SSAs in fast spectrum space reactors, preferred for compactness and higher powers, has focused on rhenium as the primary SSA. In addition to identifying additional SSAs, the present work investigates the effects of SSAs on the overall size and mass, temperature reactivity feedback, and operational lifetime of fast spectrum space reactors. The fast spectrum S4 reactor has a sectored Mo-14%Re solid-core, loadedwith UN fuel, cooled by He-30%Xe, and designed to avoid single point failures at a steady thermal power of 550 kWth. The addition of SSAs to the reactor core increases the fuel enrichment and decreases the size and mass of the reactor and the radiation shadow shield. SSA additions of boron-10, europium-151, gadolinium-155 and iridium result in the smallest and lightest S4 reactors. The effects of SSA additions on the operational lifetime and the temperature and burnup reactivity coefficients of the S^4 reactor are studied. An increasein fuel enrichment with SSAs markedly increases the operational lifetime by decreasing the burnup reactivity coefficient with only a slight decrease in the temperature reactivity feedback coefficient. With no SSAs, the UN fuel enrichment is lowest (58.5 wt%), the temperature and burnup reactivity coefficients are the highest (-0.2709 ¢/K and -1.3470 /atom%), and the estimated operating lifetime is the shortest (7.6 years). The temperature and burnup reactivity coefficients decrease to -0.2649 ¢/K and -1.0230 /atom%, and

  8. Operator Serves as Integral Member of Plant Design Team

    ERIC Educational Resources Information Center

    Norris, Dan P.; Collins, Floyd W.

    1978-01-01

    It is suggested that plant operators can be useful in designing sewage treatment plants. The advantages of this cooperative arrangement to the consulting engineers and the city, and the pitfalls, are discussed. (BB)

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

    SciTech Connect

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

    1995-07-01

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

  10. Neutronic design of a Liquid Salt-cooled Pebble Bed Reactor (LSPBR)

    SciTech Connect

    De Zwaan, S. J.; Boer, B.; Lathouwers, D.; Kloosterman, J. L.

    2006-07-01

    A renewed interest has been raised for liquid salt cooled nuclear reactors. The excellent heat transfer properties of liquid salt coolants provide several benefits, like lower fuel temperatures, higher coolant outlet temperatures, increased core power density and better decay heat removal. In order to benefit from the online refueling capability of a pebble bed reactor, the Liquid Salt Pebble Bed Reactor (LSPBR) is proposed. This is a high temperature pebble-bed reactor with a fuel design similar to existing HTRs, but using a liquid salt as a coolant. In this paper, the selection criteria for the liquid salt coolant are described. Based on its neutronic properties, LiF-BeF{sub 2} (FLIBE) was selected for the LSPBR. Two designs of the LSPBR were considered: a cylindrical core and an annular core with a graphite inner reflector. Coupled neutronic-thermal hydraulic calculations were performed to obtain the steady state power distribution and the corresponding fuel temperatures. Finally, calculations were performed to investigate the decay heat removal capability in a protected loss-of-forced cooling accident. The maximum allowable power that can be produced with the LSPBR is hereby determined. (authors)

  11. Design and operation of a rotating drum radio frequency plasma reactor for the modification of free nanoparticles.

    PubMed

    Shearer, Jeffrey C; Fisher, Ellen R

    2013-06-01

    A rotating drum rf plasma reactor was designed to functionalize the surface of nanoparticles and other unusually shaped substrates through plasma polymerization and surface modification. This proof-of-concept reactor design utilizes plasma polymerized allyl alcohol to add OH functionality to Fe2O3 nanoparticles. The reactor design is adaptable to current plasma hardware, eliminating the need for an independent reactor setup. Plasma polymerization performed on Si wafers, Fe2O3 nanoparticles supported on Si wafers, and freely rotating Fe2O3 nanoparticles demonstrated the utility of the reactor for a multitude of processes. X-ray photoelectron spectroscopy and Fourier transform infrared spectroscopy were used to characterize the surface of the substrates prior to and after plasma deposition, and scanning electron microscopy was used to verify that no extensive change in the size or shape of the nanoparticles occurred because of the rotating motion of the reactor. The reactor design was also extended to a non-depositing NH3 plasma modification system to demonstrate the reactor design is effective for multiple plasma processes.

  12. Development of a low capital investment reactor system: application for plant cell suspension culture

    PubMed

    Hsiao; Bacani; Carvalho; Curtis

    1999-01-01

    Growth of plant cell cultures is demonstrated in an uncontrolled, simple, and inexpensive plastic-lined vessel. Sustained specific growth rates of 0.22 day-1 for Hyoscyamus muticus cell suspension cultures are achieved in a low-cost gas-sparged bioreactor configuration (6.5 L working volume, wv) which is comparable to an "optimized" 5 L wv mechanically agitated fermentor. In an effort to reduce bioreactor costs, the need for an autoclavable vessel was eliminated. Sterilization is achieved by separate autoclaving of the plastic liner and by gas-phase sterilization using ethylene oxide. The initial run sterilized with ethylene oxide displayed a long lag, apparently due to residual sterilant gas. Because ethylene oxide could eliminate costs associated with autoclave rated vessels, a quantitative basis for aeration time was developed by experimental measurements and modeling of diffusion in the polymer liner. Operational techniques to eliminate toxicity are implemented to grow 0.2 kg dry weight of plant cells in 13 days in a 40 L (28.5 L wv) air-lift bioreactor without autoclave sterilization. The biomass yields for all reactors were statistically indistinguishable from shake flask culture.

  13. 850/sup 0/C VHTR plant technical description

    SciTech Connect

    Not Available

    1980-06-01

    This report describes the conceptual design of an 842-MW(t) process heat very high temperature reactor (VHTR) plant having a core outlet temperature of 850/sup 0/C (1562/sup 0/F). The reactor is a variation of the high-temperature gas-cooled reactor (HTGR) power plant concept. The report includes a description of the nuclear heat source (NHS) and of the balance of reactor plant (BORP) requirements. The design of the associated chemical process plant is not covered in this report. The reactor design is similar to a previously reported VHTR design having a 950/sup 0/C (1742/sup 0/F) core outlet temperature.

  14. Remediation of dinitrotoluene contaminated soils from former ammunition plants: soil washing efficiency and effective process monitoring in bioslurry reactors.

    PubMed

    Zhang, C; Daprato, R C; Nishino, S F; Spain, J C; Hughes, J B

    2001-10-12

    A pilot-scale bioslurry system was used to test the treatment of soils highly contaminated with 2,4-dinitrotoluene (2,4-DNT) and 2,6-dinitrotoluene (2,6-DNT). The treatment scheme involved a soil-washing process followed by two sequential aerobic slurry reactors augmented with 2,4-DNT- and 2,6-DNT-mineralizing bacteria. Test soils were obtained from two former army ammunition plants, the Volunteer Army Ammunition Plant (VAAP, Chattanooga, TN) and the Badger Army Ammunition Plant (BAAP, Baraboo, WI). Soil washing was used to minimize operational problems in slurry reactors associated with large particulates. The Eimco slurry reactors were operated in a draw-and-fill mode for 3 months and were monitored for the biodegradation of 2,4-DNT and 2,6-DNT, nitrite production, NaOH consumption, and oxygen uptake rate. Results show that soil washing was very effective for the removal of sands and the recovery of soil fines containing 2,4-DNT and 2,6-DNT. Bioslurry reactors offered rapid and nearly complete degradation of both DNT isomers, but require real time monitoring to avoid long lag periods upon refeeding. Results found a significant discrepancy between the measured DNT concentrations and calculated DNT concentrations in the slurry reactors because of solids profiles in the slurry reactors and the presence of floating crystal of DNTs. Based on the actual amount of dinitrotoluene degradation, nitrite release, NaOH consumption, and oxygen uptake were close to the theoretical stoichiometric coefficients of complete DNT mineralization. Such stoichiometric relationships were not achieved if the calculation was based on the measured DNT concentrations due to the heterogeneity of DNT in the reactor. Results indicate that nitrite release, NaOH consumption, and oxygen uptake rates provide a fast assessment of 2,4-DNT degradation and microbial activity in a slurry reactor, but could not be extended to a second reactor in series where the degradation of a much lower concentration

  15. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    SciTech Connect

    Nash, C.A.; Blake, J.E.; Rush, G.C.

    1990-12-31

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m{sup 2}) (1.1E+6 BTU/(ft{sup 2}hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient.

  16. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    SciTech Connect

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.

    1986-06-01

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost.

  17. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    SciTech Connect

    Nash, C.A. ); Blake, J.E.; Rush, G.C. )

    1990-01-01

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m{sup 2}) (1.1E+6 BTU/(ft{sup 2}hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient.

  18. Mechanical and thermal design of the Cascade reactor

    SciTech Connect

    Pitts, J.H.

    1983-01-01

    We present an improved Cascade fusion reaction chamber that is optimized with respect to chamber radius, wall thickness, and pebble blanket outlet temperature. We show results of a parameter study where we varied chamber radius from 3 to 6 m, wall thickness from 15 to 80 mm, and blanket outlet temperature from 900 to 1400 K. Based on these studies, we achieved an optimized chamber with 50% the volume of the original design and 60% of its blanket. Chamber radius is only 4.4 m and its half length is only 5.9 m, decreased from the original 5-m radius and 8-m half-length. In our optimization method, we calculate both thermal and mechanical stresses resulting from x-ray, fusion-pellet-debris, and neutron-generated momentum, pressure from ablated material, centrifugal action, vacuum inside the chamber, and gravity. We add the mechanical stresses to thermal stresses and keep the total less than the yield stress. Further, we require that fluctuations in these stresses be less than that which would produce creep-fatigue failure within the chamber 30-year lifetime.

  19. ENGINEERING ASPECTS OF COLLEGE PLANT DESIGN.

    ERIC Educational Resources Information Center

    DALTON, LIAM F.; SEGNER, MARVIN

    THE ARTICLE FOCUSES ON MECHANICAL AND ELECTRICAL FACILITIES THAT SHOULD BE CONSIDERED WHEN DEVELOPING A LONG RANGE MASTER PLAN. DEVELOPMENT OF THE MASTER PLAN SHOULD CONSIDER THE FOLLOWING--(1) COMPARATIVE FUEL COSTS, (2) POWER DISTRIBUTION, (3) HEATING PLANT, (4) CENTRAL PLANT SITE, (5) COOLING PLANT, (6) WATER SUPPLY, (7) STORM DRAINAGE, (8)…

  20. Impact of mechanical- and maintenance-induced failures of main reactor coolant pump seals on plant safety

    SciTech Connect

    Azarm, M A; Boccio, J L; Mitra, S

    1985-12-01

    This document presents an investigation of the safety impact resulting from mechanical- and maintenance-induced reactor coolant pump (RCP) seal failures in nuclear power plants. A data survey of the pump seal failures for existing nuclear power plants in the US from several available sources was performed. The annual frequency of pump seal failures in a nuclear power plant was estimated based on the concept of hazard rate and dependency evaluation. The conditional probability of various sizes of leak rates given seal failures was then evaluated. The safety impact of RCP seal failures, in terms of contribution to plant core-melt frequency, was also evaluated for three nuclear power plants. For leak rates below the normal makeup capacity and the impact of plant safety were discussed qualitatively, whereas for leak rates beyond the normal make up capacity, formal PRA methodologies were applied. 22 refs., 17 figs., 19 tabs.

  1. An investigation of temperature measurement methods in nuclear power plant reactor pressure vessel annealing

    SciTech Connect

    Acton, R.U.; Gill, W.; Sais, D.J.; Schulze, D.H.; Nakos, J.T.

    1996-05-01

    The objective of this project was to provide an assessment of several methods by which the temperature of a commercial nuclear power plant reactor pressure vessel (RPV) could be measured during an annealing process. This project was a coordinated effort between DOE`s Office of Nuclear Energy, Science and Technology; DOE`s Light Water Reactor Technology Center at Sandia National Laboratories; and the Electric Power Research Institute`s Non- Destructive Evaluation Center. Ball- thermocouple probes similar to those described in NUREG/CR-5760, spring-loaded, metal- sheathed thermocouple probes, and 1778 air- suspended thermocouples were investigated in experiments that heated a section of an RPV wall to simulate a thermal annealing treatment. A parametric study of ball material, emissivity, thermal conductivity, and thermocouple function locations was conducted. Also investigated was a sheathed thermocouple failure mode known as shunting (electrical breakdown of insulation separating the thermocouple wires). Large errors were found between the temperature as measured by the probes and the true RPV wall temperature during heat-up and cool-down. At the annealing soak temperature, in this case 454{degrees}C [850`F], all sensors measured the same temperature within about {plus_minus}5% (23.6{degrees}C [42.5{degrees}F]). Because of these errors, actual RPV wall heating and cooling rates differed from those prescribed (by up to 29%). Shunting does not appear to be a problem under these conditions. The large temperature measurement errors led to the development of a thermal model that predicts the RPV wall temperature from the temperature of a ball- probe. Comparisons between the model and the experimental data for ball-probes indicate that the model could be a useful tool in predicting the actual RPV temperature based on the indicated ball- probe temperature. The model does not predict the temperature as well for the spring-loaded and air suspended probes.

  2. Application of Nuclear Energy for Seawater Desalination: Design Concepts of Nuclear Desalination Plants

    SciTech Connect

    Faibish, R.S.; Konishi, T.; Gasparini, M.

    2002-07-01

    Nuclear energy is playing an important role in electricity generation, producing 16% of the world's electricity. However, most of the world's energy consumption is in the form of heat, in which case nuclear energy could also play an important role. In particular, process heat for seawater desalination using nuclear energy has been of growing interest to some Member States of the International Atomic Energy Agency over the past two decades. This growing interest stems from increasingly acute freshwater shortages in many arid and semi-arid zones around the world. Indeed, several national and international nuclear desalination demonstration programs are already under way or being planned. Of particular interest are projects for seawater nuclear desalination plants in coastal regions, where saline feed water can serve the dual purpose of cooling water for the nuclear reactor and as feed water for the desalination plant. In principle any nuclear reactor can provide energy (low-grade heat and/or electricity), as required by desalination processes. However, there are some additional requirements to be met under specific conditions in order to introduce nuclear desalination. Technical issues include meeting more stringent safety requirements (nuclear reactors themselves and nuclear-desalination integrated complexes in particular), and performance improvement of the integrated systems. Economic competitiveness is another important factor to be considered for a broader deployment of nuclear desalination. For technical robustness and economic competitiveness a number of design variants of coupling configurations of nuclear desalination integrated plant concepts are being evaluated. This paper identifies and discusses various factors, which support the attractiveness of nuclear desalination. It further summarizes some of the key approaches recommended for nuclear desalination complex design and gives an overview of various design concepts of nuclear desalination plants, which

  3. Design for a small-scale fuel-alcohol plant

    SciTech Connect

    Berglund, G.R.; Richardson, J.G.

    1981-06-01

    The design of a small scale fuel alcohol plant 100 L/h (26.4 gal/h) of 95% (190 proof) ethanol is presented. The plant was designed and constructed using commercially available equipment. The object was to provide an energy efficiency and economical feasible reference design of a small scale fuel alcohol plant. The design requirements of the plant are presented. Each subsystem is described in detail. The systems discussed are feedstock handling and preparation; cooking and saccharification fermentation, distillation, and the automatic control system. Also discussed are test results, and costs.

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

    NASA Astrophysics Data System (ADS)

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

    2016-06-01

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

  5. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    SciTech Connect

    Lv, Quiping; Sun, Xiaodong; Chtistensen, Richard; Blue, Thomas; Yoder, Graydon; Wilson, Dane

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  6. Design and optimization of a back-flow limiter for the high performance light water reactor

    SciTech Connect

    Fischer, Kai; Laurien, Eckart; Claas, Andreas G.; Schulenberg, Thomas

    2007-07-01

    Design and Analysis of a back-flow limiter are presented, which is implemented as a safety device in the four inlet lines of the Reactor Pressure Vessel (RPV) of the High Performance Light Water Reactor (HPLWR). As a passive component, the back-flow limiter has no moving parts and belongs to the group of fluid diodes. It has low flow resistance for regular operation condition and a high flow resistance when the flow direction is reversed which is the case if a break of the feedwater line occurs. The increased flow resistance is due to a substantially increased swirl for reverse flow condition. The design is optimized employing 1D flow analyses in combination with 3D CFD analyses with respect to geometrical modifications, like the nozzle shape and swirler angles. (authors)

  7. Design and implementation of a novel portable atomic layer deposition/chemical vapor deposition hybrid reactor.

    PubMed

    Selvaraj, Sathees Kannan; Jursich, Gregory; Takoudis, Christos G

    2013-09-01

    We report the development of a novel portable atomic layer deposition chemical vapor deposition (ALD/CVD) hybrid reactor setup. Unique feature of this reactor is the use of ALD/CVD mode in a single portable deposition system to fabricate multi-layer thin films over a broad range from "bulk-like" multi-micrometer to nanometer atomic dimensions. The precursor delivery system and control-architecture are designed so that continuous reactant flows for CVD and cyclic pulsating flows for ALD mode are facilitated. A custom-written LabVIEW program controls the valve sequencing to allow synthesis of different kinds of film structures under either ALD or CVD mode or both. The entire reactor setup weighs less than 40 lb and has a relatively small footprint of 8 × 9 in., making it compact and easy for transportation. The reactor is tested in the ALD mode with titanium oxide (TiO2) ALD using tetrakis(diethylamino)titanium and water vapor. The resulting growth rate of 0.04 nm/cycle and purity of the films are in good agreement with literature values. The ALD/CVD hybrid mode is demonstrated with ALD of TiO2 and CVD of tin oxide (SnOx). Transmission electron microscopy images of the resulting films confirm the formation of successive distinct TiO2-ALD and SnO(x)-CVD layers.

  8. Design and implementation of a novel portable atomic layer deposition/chemical vapor deposition hybrid reactor

    NASA Astrophysics Data System (ADS)

    Selvaraj, Sathees Kannan; Jursich, Gregory; Takoudis, Christos G.

    2013-09-01

    We report the development of a novel portable atomic layer deposition chemical vapor deposition (ALD/CVD) hybrid reactor setup. Unique feature of this reactor is the use of ALD/CVD mode in a single portable deposition system to fabricate multi-layer thin films over a broad range from "bulk-like" multi-micrometer to nanometer atomic dimensions. The precursor delivery system and control-architecture are designed so that continuous reactant flows for CVD and cyclic pulsating flows for ALD mode are facilitated. A custom-written LabVIEW program controls the valve sequencing to allow synthesis of different kinds of film structures under either ALD or CVD mode or both. The entire reactor setup weighs less than 40 lb and has a relatively small footprint of 8 × 9 in., making it compact and easy for transportation. The reactor is tested in the ALD mode with titanium oxide (TiO2) ALD using tetrakis(diethylamino)titanium and water vapor. The resulting growth rate of 0.04 nm/cycle and purity of the films are in good agreement with literature values. The ALD/CVD hybrid mode is demonstrated with ALD of TiO2 and CVD of tin oxide (SnOx). Transmission electron microscopy images of the resulting films confirm the formation of successive distinct TiO2-ALD and SnOx-CVD layers.

  9. Design Study of Small Pb-Bi Cooled Modified Candle Reactors

    NASA Astrophysics Data System (ADS)

    Su'ud, Zaki; Sekimoto, H.

    2010-06-01

    In this study application of modified CANDLE burnup scheme based long life Pb-Bi Cooled Fast Reactors for small long life reactors with natural Uranium as Fuel Cycle Input has been performed. The reactor cores are subdivided into several parts with the same volume in the axial directions. The natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2, and 10 years after that it is shifted to region 3. This concept is applied to all regions, i.e. shifted the core of I'th region into I+1 region after the end of 10 years burn-up cycle. The first region 1 is filled by fresh natural uranium fuel. Compared to the previous works, in a smaller reactor core the criticality need to be considered more carefully especially at the beginning of life. As an optimized design, a core of 85 cm radius and 150 cm height with 300 MWt power are selected. This core can be operated 10 years without refueling or fuel shuffling. The average discharge burn-up is 350 GWd/ton HM.

  10. Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor

    SciTech Connect

    K. L. Davis; D. L. Knudson; J. L. Rempe; J. C. Crepeau; S. Solstad

    2015-07-01

    New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can undergo significant dimensional and physical changes during high temperature irradiations. In order to accurately predict these changes, real-time data must be obtained under prototypic irradiation conditions for model development and validation. To provide such data, researchers at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) are developing several instrumented test rigs to obtain data real-time from specimens irradiated in well-controlled pressurized water reactor (PWR) coolant conditions in the Advanced Test Reactor (ATR). This paper reports the status of INL efforts to develop and evaluate prototype test rigs that rely on Linear Variable Differential Transformers (LVDTs) in laboratory settings. Although similar LVDT-based test rigs have been deployed in lower flux Materials Testing Reactors (MTRs), this effort is unique because it relies on robust LVDTs that can withstand higher temperatures and higher fluxes than often found in other MTR irradiations. Specifically, the test rigs are designed for detecting changes in length and diameter of specimens irradiated in ATR PWR loops. Once implemented, these test rigs will provide ATR users with unique capabilities that are sorely needed to obtain measurements such as elongation caused by thermal expansion and/or creep loading and diameter changes associated with fuel and cladding swelling, pellet-clad interaction, and crud buildup.

  11. Design and Performance Validation of a Conductively Heated Sealed-Vessel Reactor for Organic Synthesis.

    PubMed

    Obermayer, David; Znidar, Desiree; Glotz, Gabriel; Stadler, Alexander; Dallinger, Doris; Kappe, C Oliver

    2016-12-02

    A newly designed robust and safe laboratory scale reactor for syntheses under sealed-vessel conditions at 250 °C maximum temperature and 20 bar maximum pressure is presented. The reactor employs conductive heating of a sealed glass vessel via a stainless steel heating jacket and implements both online temperature and pressure monitoring in addition to magnetic stirring. Reactions are performed in 10 mL borosilicate vials that are sealed with a silicone cap and Teflon septum and allow syntheses to be performed on a 2-6 mL scale. This conductively heated reactor is compared to a standard single-mode sealed-vessel microwave instrument with respect to heating and cooling performance, stirring efficiency, and temperature and pressure control. Importantly, comparison of the reaction outcome for a number of different synthetic transformations performed side by side in the new device and a standard microwave reactor suggest that results obtained using microwave conditions can be readily mimicked in the operationally much simpler and smaller conventionally heated device.

  12. Design of a Solar Reactor to Split CO2 Via Isothermal Redox Cycling of Ceria

    SciTech Connect

    Bader, R; Chandran, RB; Venstrom, LJ; Sedler, SJ; Krenzke, PT; De Smith, RM; Banerjee, A; Chase, TR; Davidson, JH; Lipinski, W

    2014-12-23

    The design procedure for a 3 kWth prototype solar thermochemical reactor to implement isothermal redox cycling of ceria for CO2 splitting is presented. The reactor uses beds of mm-sized porous ceria particles contained in the annulus of concentric alumina tube assemblies that line the cylindrical wall of a solar cavity receiver. The porous particle beds provide high surface area for the heterogeneous reactions, rapid heat and mass transfer, and low pressure drop. Redox cycling is accomplished by alternating flows of inert sweep gas and CO2 through the bed. The gas flow rates and cycle step durations are selected by scaling the results from small-scale experiments. Thermal and thermo-mechanical models of the reactor and reactive element tubes are developed to predict the steady-state temperature and stress distributions for nominal operating conditions. The simulation results indicate that the target temperature of 1773K will be reached in the prototype reactor and that the Mohr-Coulomb static factor of safety is above two everywhere in the tubes, indicating that thermo-mechanical stresses in the tubes remain acceptably low.

  13. Secondary Heat Exchanger Design and Comparison for Advanced High Temperature Reactor

    SciTech Connect

    Piyush Sabharwall; Ali Siahpush; Michael McKellar; Michael Patterson; Eung Soo Kim

    2012-06-01

    The goals of next generation nuclear reactors, such as the high temperature gas-cooled reactor and advance high temperature reactor (AHTR), are to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. The need for efficiency, compactness, and safety challenge the boundaries of existing heat exchanger technology, giving rise to the following study. Various studies have been performed in attempts to update the secondary heat exchanger that is downstream of the primary heat exchanger, mostly because its performance is strongly tied to the ability to employ more efficient conversion cycles, such as the Rankine super critical and subcritical cycles. This study considers two different types of heat exchangers—helical coiled heat exchanger and printed circuit heat exchanger—as possible options for the AHTR secondary heat exchangers with the following three different options: (1) A single heat exchanger transfers all the heat (3,400 MW(t)) from the intermediate heat transfer loop to the power conversion system or process plants; (2) Two heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants, each exchanger transfers 1,700 MW(t) with a parallel configuration; and (3) Three heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants. Each heat exchanger transfers 1,130 MW(t) with a parallel configuration. A preliminary cost comparison will be provided for all different cases along with challenges and recommendations.

  14. Design, fabrication, operation and Aspen simulation of oil shale pyrolysis and biomass gasification process using a moving bed downdraft reactor

    NASA Astrophysics Data System (ADS)

    Golpour, Hassan

    Energy is the major facilitator of the modern life. Every developed and developing economy requires access to advanced sources of energy to support its growth and prosperity. Declining worldwide crude oil reserves and increasing energy needs has focused attention on developing existing unconventional fossil fuels like oil shale and renewable resources such as biomass. Sustainable, renewable and reliable resources of domestically produced biomass comparing to wind and solar energy is a sensible motivation to establish a small-scale power plant using biomass as feed to supply electricity demand and heat for rural development. The work in Paper I focuses on the possibility of water pollution from spent oil shale which should be studied before any significant commercial production is attempted. In Paper II, the proposed Aspen models for oil shale pyrolysis is to identify the key process parameters for the reactor and optimize the rate of production of syncrude from oil shale. The work in Paper III focuses on (1) Design and operation of a vertical downdraft reactor, (2) Establishing an optimum operating methodology and parameters to maximize syngas production through process testing. Finally in Paper IV, a proposed Aspen model for biomass gasification simulates a real biomass gasification system discussed in Paper III.

  15. CORE ANALYSIS, DESIGN AND OPTIMIZATION OF A DEEP-BURN PEBBLE BED REACTOR

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-05-01

    Achieving a high burnup in the Deep-Burn pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. To investigate the aforementioned effects a code system using existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

  16. Combining COMSOL modeling with acoustic pressure maps to design sono-reactors.

    PubMed

    Wei, Zongsu; Weavers, Linda K

    2016-07-01

    Scaled-up and economically viable sonochemical systems are critical for increased use of ultrasound in environmental and chemical processing applications. In this study, computational simulations and acoustic pressure maps were used to design a larger-scale sono-reactor containing a multi-stepped ultrasonic horn. Simulations in COMSOL Multiphysics showed ultrasonic waves emitted from the horn neck and tip, generating multiple regions of high acoustic pressure. The volume of these regions surrounding the horn neck were larger compared with those below the horn tip. The simulated acoustic field was verified by acoustic pressure contour maps generated from hydrophone measurements in a plexiglass box filled with water. These acoustic pressure contour maps revealed an asymmetric and discrete distribution of acoustic pressure due to acoustic cavitation, wave interaction, and water movement by ultrasonic irradiation. The acoustic pressure contour maps were consistent with simulation results in terms of the effective scale of cavitation zones (∼ 10 cm and <5 cm above and below horn tip, respectively). With the mapped acoustic field and identified cavitation location, a cylindrically-shaped sono-reactor with a conical bottom was designed to evaluate the treatment capacity (∼ 5 L) for the multi-stepped horn using COMSOL simulations. In this study, verification of simulation results with experiments demonstrates that coupling of COMSOL simulations with hydrophone measurements is a simple, effective and reliable scientific method to evaluate reactor designs of ultrasonic systems.

  17. Radioactive waste treatment in nuclear power plants with VVER reactors using Balakovo NPP (Russia) as an example

    SciTech Connect

    Kutscher, U.; Hoelker, G.; Chrubasik, A.; Ipatov, P.L.; Ignatov, V.I.

    1993-12-31

    In nuclear power plants with VVER reactors, located in the states of the former USSR, the treatment of solid and liquid radioactive waste represents a problem which--besides measures concerning the increase of reactor safety--requires a quick solution at a high technical level. To help address this problem, Nukem GmbH jointly with Balakovo NPP and the Nuclear Power Project Institute AEP Nizhni Novgorod, has developed a waste Treatment Center (WTC) for the Balakowo NPP. This center possesses all the necessary technological installations to transform liquid low-level, medium-level and solid low-level radioactive waste accumulated to date and waste arising during the futures operation of Balakovo NPP (4 reactors VVER-1000) in a way which allows a safe long-time storage.

  18. Clinch River breeder reactor sodium fire protection system design and development

    SciTech Connect

    Foster, K.W.; Boasso, C.J.; Kaushal, N.N.

    1984-04-13

    To assure the protection of the public and plant equipment, improbable accidents were hypothesized to form the basis for the design of safety systems. One such accident is the postulated failure of the Intermediate Heat Transfer System (IHTS) piping within the Steam Generator Building (SGB), resulting in a large-scale sodium fire. This paper discusses the design and development of plant features to reduce the consequences of the accident to acceptable levels. Additional design solutions were made to mitigate the sodium spray contribution to the accident scenario. Sodium spill tests demonstrated that large sodium leaks can be safely controlled in a sodium-cooled nuclear power plant.

  19. Preliminary design of the Carrisa Plains solar central receiver power plant. Volume II. Plant specifications

    SciTech Connect

    Price, R. E.

    1983-12-31

    The specifications and design criteria for all plant systems and subsystems used in developing the preliminary design of Carrisa Plains 30-MWe Solar Plant are contained in this volume. The specifications have been organized according to plant systems and levels. The levels are arranged in tiers. Starting at the top tier and proceeding down, the specification levels are the plant, system, subsystem, components, and fabrication. A tab number, listed in the index, has been assigned each document to facilitate document location.

  20. H-Coal process and plant design

    DOEpatents

    Kydd, Paul H.; Chervenak, Michael C.; DeVaux, George R.

    1983-01-01

    A process for converting coal and other hydrocarbonaceous materials into useful and more valuable liquid products. The process comprises: feeding coal and/or other hydrocarbonaceous materials with a hydrogen-containing gas into an ebullated catalyst bed reactor; passing the reaction products from the reactor to a hot separator where the vaporous and distillate products are separated from the residuals; introducing the vaporous and distillate products from the separator directly into a hydrotreater where they are further hydrogenated; passing the residuals from the separator successively through flash vessels at reduced pressures where distillates are flashed off and combined with the vaporous and distillate products to be hydrogenated; transferring the unseparated residuals to a solids concentrating and removal means to remove a substantial portion of solids therefrom and recycling the remaining residual oil to the reactor; and passing the hydrogenated vaporous and distillate products to an atmospheric fractionator where the combined products are fractionated into separate valuable liquid products. The hydrogen-containing gas is generated from sources within the process.

  1. Evaluation of a Business Case for Safeguards by Design in Nuclear Power Reactors

    SciTech Connect

    Wood, Thomas W.; Seward, Amy M.; Lewis, Valerie A.; Gitau, Ernest TN; Zentner, Michael D.

    2012-12-01

    Safeguards by Design (SbD) is a well-known paradigm for consideration and incorporation of safeguards approaches and associated design features early in the nuclear facility development process. This paradigm has been developed as part of the Next Generation Safeguards Initiative (NGSI), and has been accepted as beneficial in many discussions and papers on NGSI or specific technologies under development within NGSI. The Office of Nuclear Safeguards and Security funded the Pacific Northwest National Laboratory to examine the business case justification of SbD for nuclear power reactors. Ultimately, the implementation of SbD will rely on the designers of nuclear facilities. Therefore, it is important to assess the incentives which will lead designers to adopt SbD as a standard practice for nuclear facility design. This report details the extent to which designers will have compelling economic incentives to adopt SbD.

  2. Safety design approach for external events in Japan sodium-cooled fast reactor

    SciTech Connect

    Yamano, H.; Kubo, S.; Tani, A.; Nishino, H.; Sakai, T.

    2012-07-01

    This paper describes a safety design approach for external events in the design study of Japan sodium-cooled fast reactor. An emphasis is introduction of a design extension external condition (DEEC). In addition to seismic design, other external events such as tsunami, strong wind, abnormal temperature, etc. were addressed in this study. From a wide variety of external events consisting of natural hazards and human-induced ones, a screening method was developed in terms of siting, consequence, frequency to select representative events. Design approaches for these events were categorized on the probabilistic, statistical and deterministic basis. External hazard conditions were considered mainly for DEECs. In the probabilistic approach, the DEECs of earthquake, tsunami and strong wind were defined as 1/10 of exceedance probability of the external design bases. The other representative DEECs were also defined based on statistical or deterministic approaches. (authors)

  3. A UASB reactor coupled to a hybrid aerobic MBR as innovative plant configuration to enhance the removal of organic micropollutants.

    PubMed

    Alvarino, T; Suárez, S; Garrido, M; Lema, J M; Omil, F

    2016-02-01

    An innovative plant configuration based in an UASB reactor coupled to a hybrid aerobic membrane bioreactor designed for sustainable treatment of municipal wastewater at ambient temperatures and low hydraulic retention time was studied in terms of organic micropollutants (OMPs) removal. OMPs removal mechanisms, as well as the potential influence of biomass activity and physical conformation were assessed. Throughout all periods of operation (150 days) high organic matter removals were maintained (>95%) and, regarding OMPs removal, this innovative system has shown to be more efficient than conventional technologies for those OMPs which are prone to be biotransformed under anaerobic conditions. For instance, sulfamethoxazole and trimethoprim have both shown to be biodegradable under anaerobic conditions with similar efficiencies (removal efficiencies above 84%). OMPs main removal mechanism was found to be biotransformation, except in the case of musk fragrances which showed medium sorption onto sludge. OMPs removal was strongly dependent on the efficiency of the primary metabolism (organic matter degradation and nitrification) and the type of biomass.

  4. Demonstration of a 100-kWth high-temperature solar thermochemical reactor pilot plant for ZnO dissociation

    NASA Astrophysics Data System (ADS)

    Koepf, E.; Villasmil, W.; Meier, A.

    2016-05-01

    Solar thermochemical H2O and CO2 splitting is a viable pathway towards sustainable and large-scale production of synthetic fuels. A reactor pilot plant for the solar-driven thermal dissociation of ZnO into metallic Zn has been successfully developed at the Paul Scherrer Institute (PSI). Promising experimental results from the 100-kWth ZnO pilot plant were obtained in 2014 during two prolonged experimental campaigns in a high flux solar simulator at PSI and a 1-MW solar furnace in Odeillo, France. Between March and June the pilot plant was mounted in the solar simulator and in-situ flow-visualization experiments were conducted in order to prevent particle-laden fluid flows near the window from attenuating transparency by blocking incoming radiation. Window flow patterns were successfully characterized, and it was demonstrated that particle transport could be controlled and suppressed completely. These results enabled the successful operation of the reactor between August and October when on-sun experiments were conducted in the solar furnace in order to demonstrate the pilot plant technology and characterize its performance. The reactor was operated for over 97 hours at temperatures as high as 2064 K; over 28 kg of ZnO was dissociated at reaction rates as high as 28 g/min.

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

    SciTech Connect

    Not Available

    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.

  6. 45 CFR 670.21 - Designation of native plants.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 45 Public Welfare 3 2011-10-01 2011-10-01 false Designation of native plants. 670.21 Section 670.21 Public Welfare Regulations Relating to Public Welfare (Continued) NATIONAL SCIENCE FOUNDATION CONSERVATION OF ANTARCTIC ANIMALS AND PLANTS Native Mammals, Birds, Plants, and Invertebrates §...

  7. 45 CFR 670.21 - Designation of native plants.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 45 Public Welfare 3 2012-10-01 2012-10-01 false Designation of native plants. 670.21 Section 670.21 Public Welfare Regulations Relating to Public Welfare (Continued) NATIONAL SCIENCE FOUNDATION CONSERVATION OF ANTARCTIC ANIMALS AND PLANTS Native Mammals, Birds, Plants, and Invertebrates §...

  8. 45 CFR 670.21 - Designation of native plants.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 45 Public Welfare 3 2010-10-01 2010-10-01 false Designation of native plants. 670.21 Section 670.21 Public Welfare Regulations Relating to Public Welfare (Continued) NATIONAL SCIENCE FOUNDATION CONSERVATION OF ANTARCTIC ANIMALS AND PLANTS Native Mammals, Birds, Plants, and Invertebrates §...

  9. 45 CFR 670.21 - Designation of native plants.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 45 Public Welfare 3 2014-10-01 2014-10-01 false Designation of native plants. 670.21 Section 670.21 Public Welfare Regulations Relating to Public Welfare (Continued) NATIONAL SCIENCE FOUNDATION CONSERVATION OF ANTARCTIC ANIMALS AND PLANTS Native Mammals, Birds, Plants, and Invertebrates §...

  10. 45 CFR 670.21 - Designation of native plants.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 45 Public Welfare 3 2013-10-01 2013-10-01 false Designation of native plants. 670.21 Section 670.21 Public Welfare Regulations Relating to Public Welfare (Continued) NATIONAL SCIENCE FOUNDATION CONSERVATION OF ANTARCTIC ANIMALS AND PLANTS Native Mammals, Birds, Plants, and Invertebrates §...

  11. Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters

    SciTech Connect

    Darmann, Frank; Lombaerde, Robert; Moriconi, Franco; Nelson, Albert

    2012-03-01

    Zenergy Power has successfully designed, built, tested, and installed in the US electrical grid a saturable reactor Fault Current Limiter. Beginning in 2007, first as SC Power Systems and from 2008 as Zenergy Power, Inc., ZP used DOE matching grant and ARRA funds to help refine the design of the saturated reactor fault current limiter. ZP ultimately perfected the design of the saturated reactor FCL to the point that ZP could reliably design a suitable FCL for most utility applications. Beginning with a very basic FCL design using 1G HTS for a coil housed in a LN2 cryostat for the DC bias magnet, the technology progressed to a commercial system that was offered for sale internationally. Substantial progress was made in two areas. First, the cryogenics cooling system progressed from a sub-cooled liquid nitrogen container housing the HTS coils to cryostats utilizing dry conduction cooling and reaching temperatures down to less than 20 degrees K. Large, round cryostats with warm bore diameters of 1.7 meters enabled the design of large tanks to hold the AC components. Second, the design of the AC part of the FCL was refined from a six legged spider design to a more compact and lighter design with better fault current limiting capability. Further refinement of the flux path and core shape led to an efficient saturated reactor design requiring less Ampere-turns to saturate the core. In conclusion, the development of the saturable reactor FCL led to a more efficient design not requiring HTS magnets and their associated peripheral equipment, which yielded a more economical product in line with the electric utility industry expectations. The original goal for the DOE funding of the ZP project Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters was to stimulate the HTS wire industry with, first 1G, then 2G, HTS wire applications. Over the approximately 5 years of ZP's product development program, the amount of HTS wire

  12. 1170 MW/sub t/ HTGR steamer cogeneration plant: design and cost study

    SciTech Connect

    1980-08-01

    A conceptual design and cost study is presented for intermediate size high temperature gas-cooled reactor (HTGR) for industrial energy applications performed by United Engineers and Constructors Inc., (UE and C) and The General Atomic Company (GAC). The study is part of a program at ORNL and has the objective to provide support in the evaluation of the technical and economic feasibility of a single unit 1170 MW/sub t/ HTGR steam cycle cogeneration plant (referred to as the Steamer plant) for the production of industrial process energy. Inherent in the achievement of this objective, it was essential to perform a number of basic tasks such as the development of plant concept, capital cost estimate, project schedule and annual operation and maintenance (O and M) cost.

  13. Design and analysis of megawatt-class heat-pipe reactor concepts

    SciTech Connect

    Poston, D.; Kapernick, R.

    2012-07-01

    There is growing interest in finding an alternative to diesel-powered systems at locations removed from a reliable electrical grid. One promising option is a 1- to 10-MW mobile reactor system, that could provide robust, self-contained, and long-term ({>=} 5 years) power in any environment. The reactor and required infrastructure could be transported to any location within one or a few standard transport containers. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than 'traditional' reactors that rely on pumped coolant through the core. This paper examines a heat pipe reactor that is fabricated and shipped as six identical core segments. Each core segment includes a heat-pipe-to-gas heat exchanger that is coupled to the condenser end of the heat pipes. The reference power conversion system is a CO{sub 2}-Brayton system. The segments by themselves are deeply subcritical during transport, and they would be locked into an operating configuration (with control inserted) at the final destination. Two design options are considered: a near-term option and an advanced option. The near-term option is a 5-MWt concept that uses uranium-dioxide fuel, a stainless-steel structure, and potassium as the heat-pipe working fluid. The advanced option is a 15-MWt concept that uses uranium-nitride fuel, a molybdenum/TZM structure, and sodium as the heat-pipe working fluid. The materials used in the advanced option allow for higher temperatures and power densities, and enhanced power throughput in the heat pipes. Higher powers can be obtained from both concepts by increasing the core size and the number of heat pipes. (authors)

  14. A Fusion Reactor Design with a Liquid First Wall and Divertor

    SciTech Connect

    Nygren, R E; Rognlien, T D; Rensink, M E; Smolentsev, S S; Youssef, M E; Sawan, M Z; Merrill, B J; Eberle, C; Fogarty, P J; Nelson, B E; Sze, D K; Majeski, R

    2003-11-13

    Within the magnetic fusion energy program in the US, a program called APEX is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around the plasma. As part of this work, the APEX Team has investigated several possible design implementations and developed a specific engineering concept for a fusion reactor with liquid walls. Our approach has been to utilize an already established design for a future fusion reactor, the ARIES-RS, for the basic chamber geometry and magnetic configuration and to replace the chamber technology in this design with liquid wall technology for a first wall and divertor and a blanket with adequate tritium breeding. This paper gives an overview of one design with a molten salt (a mixture of lithium, beryllium and sodium fluorides) forming the liquid surfaces and a ferritic steel for the structural material of the blanket. The design point is a reactor with 3840MW of fusion power of which 767MW is in the form of energetic particles (alpha power) and 3073MW is in the form of neutrons. The alpha plus auxiliary power total 909MW of which 430MW is radiated from the core mostly onto the first wall and the balance flows into the edge plasma and is distributed between the first wall and the divertor. In pursuing the application of liquid surfaces in APEX, the team has developed analytical tools that are significant achievements themselves and also pursued experiments on flowing liquids. This work is covered elsewhere, but the paper will also note several such areas to indicate the supporting science behind the design presented. Significant new work in modeling the plasma edge to understand the interaction of the plasma with the liquid walls is one example. Another is the incorporation of magneto-hydrodynamic (MHD) effects in fluid modeling and heat transfer.

  15. TF Inner Leg Space Allocation for Pilot Plant Design Studies

    SciTech Connect

    Peter H. Titus and Ali Zolfaghari

    2012-09-06

    A critical design feature of any tokamak is the space taken up by the inner leg of the toroidal field (TF) coil. The radial build needed for the TF inner leg, along with shield thickness , size of the central solenoid and plasma minor radius set the major radius of the machine. The cost of the tokamak core roughly scales with the cube of the major radius. Small reductions in the TF build can have a big impact on the overall cost of the reactor. The cross section of the TF inner leg must structurally support the centering force and that portion of the vertical separating force that is not supported by the outer structures. In this paper, the TF inner leg equatorial plane cross sections are considered. Out-of- Plane (OOP) forces must also be supported, but these are largest away from the equatorial plane, in the inner upper and lower corners and outboard sections of the TF coil. OOP forces are taken by structures that are not closely coupled with the radial build of the central column at the equatorial plane. The "Vertical Access AT Pilot Plant" currently under consideration at PPPL is used as a starting point for the structural, field and current requirements. Other TF structural concepts are considered. Most are drawn from existing designs such as ITER's circular conduits in radial plates bearing on a heavy nose section, and TPX's square conduits in a case, Each of these concepts can rely on full wedging, or partial wedging. Vaulted TF coils are considered as are those with some component of bucking against a central solenoid or bucking post. With the expectation that the pilot plant will be a steady state machine, a static stress criteria is used for all the concepts. The coils are assumed to be superconducting, with the superconductor not contributing to the structural strength. Limit analysis is employed to assess the degree of conservatism in the static criteria as it is applied to a linear elastic stress analysis. TF concepts, and in particular the PPPL AT

  16. Design of Biomass Gasification and Combined Heat and Power Plant Based on Laboratory Experiments

    NASA Astrophysics Data System (ADS)

    Haydary, Juma; Jelemenský, Ľudovít

    Three types of wooden biomass were characterized by calorimetric measurements, proximate and elemental analysis, thermogravimetry, kinetics of thermal decomposition and gas composition. Using the Aspen steady state simulation, a plant with the processing capacity of 18 ton/h of biomass was modelled based on the experimental data obtained under laboratory conditions. The gasification process has been modelled in two steps. The first step of the model describes the thermal decomposition of the biomass based on a kinetic model and in the second step, the equilibrium composition of syngas is calculated by the Gibbs free energy of the expected components. The computer model of the plant besides the reactor model includes also a simulation of other plant facilities such as: feed drying employing the energy from the process, ash and tar separation, gas-steam cycle, and hot water production heat exchangers. The effect of the steam to air ratio on the conversion, syngas composition, and reactor temperature was analyzed. Employment of oxygen and air for partial combustion was compared. The designed computer model using all Aspen simulation facilities can be applied to study different aspects of biomass gasification in a Combined Heat and Power plant.

  17. 21 CFR 129.20 - Plant construction and design.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 21 Food and Drugs 2 2013-04-01 2013-04-01 false Plant construction and design. 129.20 Section 129.20 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED... § 129.20 Plant construction and design. (a) The bottling room shall be separated from other...

  18. 21 CFR 129.20 - Plant construction and design.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 21 Food and Drugs 2 2014-04-01 2014-04-01 false Plant construction and design. 129.20 Section 129.20 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED... § 129.20 Plant construction and design. (a) The bottling room shall be separated from other...

  19. 21 CFR 129.20 - Plant construction and design.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 21 Food and Drugs 2 2012-04-01 2012-04-01 false Plant construction and design. 129.20 Section 129.20 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED... § 129.20 Plant construction and design. (a) The bottling room shall be separated from other...

  20. Plant Design for the Production of DUAGG

    SciTech Connect

    Ferrada, J.J.

    2003-02-04

    The cost of producing DUAGG is an important consideration for any interested private firm in determining whether DUCRETE is economically viable as a material of construction in next-generation spent nuclear fuel casks. This study analyzed this project as if it was a stand-alone project. The capital cost includes engineering design, equipment costs and installation, start up, and management; the study is not intended to be a life-cycle cost analysis. The costs estimated by this study are shown in Table ES.1, and the conclusions of this study are listed in Table ES.2. The development of DUAGG and DUCRETE is a major thrust of the Depleted Uranium Uses Research and Development Project. An obvious use of depleted uranium is as a shielding material (e.g., DUCRETE). DUCRETE is made by replacing the conventional stone aggregate in concrete with DUAGG. One objective of this project is to bring the development of DUCRETE to a point at which a demonstrated basis exists for its commercial deployment. The estimation of the costs to manufacture DUAGG is an important part of this effort. Paul Lessing and William Quapp developed DUAGG and DUCRETE as part of an Idaho National Engineering and Environmental Laboratory (INEEL) program to find beneficial uses for depleted uranium (DU). Subsequently, this technology was licensed to Teton Technologies, Inc. The DUAGG process mixes DUO{sub 2} with sintering materials and additives to form pressed briquettes. These briquettes are sintered at 1300 C, and the very dense sintered briquettes are then crushed and classified into gap-graded size fractions. The graded DUAGG is then ready to be used to make high-strength heavy DUCRETE. The DUCRETE shielding will be placed into an annular steel cask-shell mold, which has internal steel reinforcing bars. The objectives of this study are to (1) use previous DUAGG process developments to design a plant that will produce DUAGG at a baseline rate, (2) determine the size of the equipment required to meet