STEAM GENERATOR FOR GAS COOLED NUCLEAR REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1960-03-14

A steam generator for a gas-cooled nuclear reactor is disposed inside the same pressure vessel as the reactor and has a tube system heated by the gas circulating through the reactor; the pressure vessel is double-walled, and the interspace between these two walls is filled with concrete serving as radiation shielding. The steam generator has a cylindricaIly shaped vertical casing, through which the heating gas circulates, while the tubes are arranged in a plurality of parallel horizontal planes and each of them have the shape of an involute of a circle. The tubes are uniformly distributed over the available surfacemore » in the plane, all the tubes of the same plane being connected in parallel. The exterior extremities of these involute-shaped tubes are each connected with similar tubes disposed in the adjacent lower situated plane, while the interior extremities are connected with tubes in the adjacent higher situated plane. The alimentation of the tubes is performed over annular headers. The tube system is self-supporting, the tubes being joined together by welded spacers. The fluid flow in the tubes is performed by forced circulation. (NPO)« less

Analysis of steam generator tube rupture transients with single failure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trambauer, K.

The Gesellschaft fuer Reaktorsicherheit is engaged in the collection and evaluation of light water reactor operating experience as well as analyses for the risk study of the pressurized water reactor (PWR). Within these activities, thermohydraulic calculations have been performed to show the influence of different boundary conditions and disturbances on the steam generator tube rupture (SGTR) transients. The analyses of these calculations have focused on the measures and systems needed to cope with an SGTR. The reference plant for this analysis is a 1300-MW(e) PWR of Kraftwerk Union design with four loops, each containing a U-tube steam generator (SG) andmore » a reactor cooling pump (RCP). The thermal-hydraulic code DRUFAN-02 was used for the transient calculations.« less

Nuclear reactor vessel fuel thermal insulating barrier

DOEpatents

Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

2013-03-19

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

Heat dissipating nuclear reactor

DOEpatents

Hunsbedt, A.; Lazarus, J.D.

1985-11-21

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

Heat dissipating nuclear reactor

DOEpatents

Hunsbedt, Anstein; Lazarus, Jonathan D.

1987-01-01

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

The thermal circuit of a nuclear power station's unit built around a supercritical-pressure water-cooled reactor

NASA Astrophysics Data System (ADS)

Silin, V. A.; Zorin, V. M.; Tagirov, A. M.; Tregubova, O. I.; Belov, I. V.; Povarov, P. V.

2010-12-01

Main results obtained from calculations of the steam generator and thermal circuit of the steam turbine unit for a nuclear power unit with supercritical-pressure water coolant and integral layout are presented. The obtained characteristics point to the advisability of carrying out further developments of this promising nuclear power technology.

Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

NASA Astrophysics Data System (ADS)

Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

2014-08-01

The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

Steam generator support system

DOEpatents

Moldenhauer, J.E.

1987-08-25

A support system for connection to an outer surface of a J-shaped steam generator for use with a nuclear reactor or other liquid metal cooled power source is disclosed. The J-shaped steam generator is mounted with the bent portion at the bottom. An arrangement of elongated rod members provides both horizontal and vertical support for the steam generator. The rod members are interconnected to the steam generator assembly and a support structure in a manner which provides for thermal distortion of the steam generator without the transfer of bending moments to the support structure and in a like manner substantially minimizes forces being transferred between the support structure and the steam generator as a result of seismic disturbances. 4 figs.

Steam generator support system

DOEpatents

Moldenhauer, James E.

1987-01-01

A support system for connection to an outer surface of a J-shaped steam generator for use with a nuclear reactor or other liquid metal cooled power source. The J-shaped steam generator is mounted with the bent portion at the bottom. An arrangement of elongated rod members provides both horizontal and vertical support for the steam generator. The rod members are interconnected to the steam generator assembly and a support structure in a manner which provides for thermal distortion of the steam generator without the transfer of bending moments to the support structure and in a like manner substantially minimizes forces being transferred between the support structure and the steam generator as a result of seismic disturbances.

Investigation of the characteristics of a compact steam reformer integrated with a water-gas shift reactor

NASA Astrophysics Data System (ADS)

Seo, Yong-Seog; Seo, Dong-Joo; Seo, Yu-Taek; Yoon, Wang-Lai

The objective of this study is to investigate numerically a compact steam methane reforming (SMR) system integrated with a water-gas shift (WGS) reactor. Separate numerical models are established for the combustion part, SMR and WGS reaction bed. The concentration of species at the exits of the SMR and WGS bed, and the temperatures in the WGS bed are in good agreement with the measured data. Heat transfer to the catalyst beds and the catalytic reactions in the SMR and WGS catalyst bed are investigated as a function of the operation parameters. The conversion of methane at the exit of the SMR catalyst bed is calculated to be 87%, and the carbon monoxide concentration at the outlet of the WGS bed is estimated to be 0.45%. The effects of the cooling heat flux at the outside wall of the system and steam-to-carbon (S/C) ratio are also examined. As the cooling heat flux increases, both the methane conversion and carbon monoxide content are reduced in the SMR bed, and the carbon monoxide conversion is improved in the WGS bed. Both methane conversion and carbon dioxide reduction increase with increasing steam-to-carbon ratio.

ETR AND MTR COMPLEXES IN CONTEXT. CAMERA FACING NORTHERLY. FROM ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR AND MTR COMPLEXES IN CONTEXT. CAMERA FACING NORTHERLY. FROM BOTTOM TO TOP: ETR COOLING TOWER, ELECTRICAL BUILDING AND LOW-BAY SECTION OF ETR BUILDING, HEAT EXCHANGER BUILDING (WITH U SHAPED YARD), COMPRESSOR BUILDING. MTR REACTOR SERVICES BUILDING IS ATTACHED TO SOUTH WALL OF MTR. WING A IS ATTACHED TO BALCONY FLOOR OF MTR. NEAR UPPER RIGHT CORNER OF VIEW IS MTR PROCESS WATER BUILDING. WING B IS AT FAR WEST END OF COMPLEX. NEAR MAIN GATE IS GAMMA FACILITY, WITH "COLD" BUILDINGS BEYOND: RAW WATER STORAGE TANKS, STEAM PLANT, MTR COOLING TOWER PUMP HOUSE AND COOLING TOWER. INL NEGATIVE NO. 56-4101. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

STEAM PLANT, TRA609. SECTION A SHOWS FEATURES OF NORTH/SOUTH AXIS: ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

STEAM PLANT, TRA-609. SECTION A SHOWS FEATURES OF NORTH/SOUTH AXIS: STEAM GENERATOR AND CATWALK, STACK, DEGREASER FEED WATER HEATER IN PENTHOUSE, MEZZANINE, SURGE TANK PIT (BELOW GROUND LEVEL). UTILITY ROOM SHOWS DIESEL ENGINE GENERATORS, AIR TANKS, STARTING AIR COMPRESSORS. OUTSIDE SOUTH END ARE EXHAUST MUFFLER, AIR INTAKE OIL FILTER, RADIATOR COOLING UNIT, AIR SURGE TANK. SECTION B CROSSES WEST TO EAST NEAR SOUTH END OF BUILDING TO SHOW ARRANGEMENT OF DIESEL ENGINE GENERATOR, AIR DRIER, AFTER COOLER, AIR COMPRESSOR, AND BLOWDOWN TANK. BLAW-KNOX 3150-9-2, 6/1950. INL INDEX NO. 431-0609-00-098-100018, REV. 3. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Optimization of steam generators of NPP with WWER in operation with variable load

NASA Astrophysics Data System (ADS)

Parchevskii, V. M.; Shchederkina, T. E.; Gur'yanova, V. V.

2017-11-01

The report addresses the issue of the optimal water level in the horizontal steam generators of NPP with WWER. On the one hand, the level needs to be kept at the lower limit of the allowable range, as gravity separation, steam will have the least humidity and the turbine will operate with higher efficiency. On the other hand, the higher the level, the greater the supply of water in the steam generator, and therefore the higher the security level of the unit, because when accidents involving loss of cooling of the reactor core, the water in the steam generators, can be used for cooling. To quantitatively compare the damage from higher level to the benefit of improving the safety was assessed of the cost of one cubic meter of water in the steam generators, the formulated objective function of optimal levels control. This was used two-dimensional separation characteristics of steam generators. It is demonstrated that the security significantly shifts the optimal values of the levels toward the higher values, and this bias is greater the lower the load unit.

Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

1976-04-01

Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

Energy alternative for industry: the high-temperature gas-cooled reactor steamer

DOE Office of Scientific and Technical Information (OSTI.GOV)

McMain, A.T. Jr.; Blok, F.J.

1978-04-01

Large industrial complexes are faced with new requirements that will lead to a transition from such fluid fuels as natural gas and oil to such solid fuels as coal and uranium for supply of industrial energy. Power plants using these latter fuels will be of moderate size (800 to 1200 MW(thermal)) and will generally have the capability of co-generating electric power and process steam. A study has been made regarding use of the 840-MW(thermal) Fort St. Vrain high-temperature gas-cooled reactor (HTGR) design for industrial applications. The initial conceptual design (referred to as the HTGR Steamer) is substantially simplified relative tomore » Fort St. Vrain in that outlet helium and steam temperatures are lower and the reheat section is deleted from the steam generators. The Steamer has four independent steam generating loops producing a total of 277 kg/s (2.2 x 10/sup 6/ lb/h) of prime steam at 4.5 MPa/672 K (650 psia/750/sup 0/F). The unit co-generates 46 MW(electric) and provides process steam at 8.31 MPa/762 K(1200 psia/912/sup 0/F). The basic configuration and much of the equipment are retained from the Fort St. Vrain design. The system has inherent safety features important for industrial applications. These and other features indicate that the HTGR Steamer is an industrial energy option deserving additional evaluation. Subsequent work will focus on parallel design optimization and application studies.« less

Collecting and recirculating condensate in a nuclear reactor containment

DOEpatents

Schultz, Terry L.

1993-01-01

An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank.

Collecting and recirculating condensate in a nuclear reactor containment

DOEpatents

Schultz, T.L.

1993-10-19

An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank. 3 figures.

Assessment of the SRI Gasification Process for Syngas Generation with HTGR Integration -- White Paper

DOE Office of Scientific and Technical Information (OSTI.GOV)

A.M. Gandrik

2012-04-01

This white paper is intended to compare the technical and economic feasibility of syngas generation using the SRI gasification process coupled to several high-temperature gas-cooled reactors (HTGRs) with more traditional HTGR-integrated syngas generation techniques, including: (1) Gasification with high-temperature steam electrolysis (HTSE); (2) Steam methane reforming (SMR); and (3) Gasification with SMR with and without CO2 sequestration.

The pre-conceptual design of the nuclear island of ASTRID

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saez, M.; Menou, S.; Uzu, B.

The CEA is involved in a substantial effort on the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) pre-conceptual design in cooperation with EDF, as experienced Sodium-cooled Fast Reactor (SFR) operator, AREVA, as experienced SFR Nuclear Island engineering company and components designer, ALSTOM POWER as energy conversion system designer and COMEX NUCLEAIRE as mechanical systems designer. The CEA is looking for other partnerships, in France and abroad. The ASTRID preliminary design is based on a sodium-cooled pool reactor of 1500 MWth generating about 600 MWe, which is required to guarantee the representativeness of the reactor core and the main componentsmore » with regard to future commercial reactors. ASTRID lifetime target is 60 years. Two Energy Conversion Systems are studied in parallel until the end of 2012: Rankine steam cycle or Brayton gas based energy conversion cycle. ASTRID design is guided by the following major objectives: improved safety, simplification of structures, improved In Service Inspection and Repair (ISIR), improved manufacturing conditions for cost reduction and increased quality, reduction of risks related to sodium fires and water/sodium reaction, and improved robustness against external hazards. The core is supported by a diagrid, which lay on a strong back to transfer the weight to the main vessel. AREVA is involved in a substantial effort in order to improve the core support structure in particular regarding the ISIR and the connection to primary pump. In the preliminary design, the primary system is formed by the main vessel and the upper closure comprising the reactor roof, two rotating plugs - used for fuel handling - and the components plugs located in the roof penetrations. The Above Core Structure deflects the sodium flow in the hot pool and provides support to core instrumentation and guidance of the control rod drive mechanisms. The number of the major components in the main vessel, primary pumps, Intermediate Heat Exchangers, and Decay Heat Exchangers are now under consideration. Under normal conditions, power release is achieved using the steam/water plant (in case of Rankine steam cycle) or the gas plant (in case of Brayton gas cycle). The diverse design and operating modes of Decay Heat Removal systems provide protection against common cause failures. A Decay Heat Removal system through the reactor vault is in particular studied with the objective to complement Direct Reactor Cooling systems. At this stage of the studies, the secondary system comprises four independent sodium loops (two and three sodium loops configurations are also investigated). Each loop includes one mechanical pump (or a large capacity Annular Linear Induction Electromagnetic Pump), and three modular Steam Generator Units characterized by once through straight tube units with a ferritic tube bundle; nevertheless, helical coil steam generator with tubes made of Alloy 800, and inverted type steam generator with a ferritic tube bundle are also investigated. The limited power of each modular Steam Generator Unit allows the whole secondary loop to withstand a large water/sodium reaction consecutive to the postulated simultaneous rupture of all the heat exchange tubes of one module. The arrangement of the components is based on the 'Regain' concept, in which the secondary pump is situated at a low level in the circuit; conventional arrangement, as SUPERPHENIX type, is a back-up option. Alternative arrangements based on gas cycles are also studied together with Na-gas heat exchanger design. This paper presents a status of the ASTRID pre-conceptual design. The most promising options are highlighted as well as less risky and back-up options. (authors)« less

Passive containment cooling system with drywell pressure regulation for boiling water reactor

DOEpatents

Hill, Paul R.

1994-01-01

A boiling water reactor having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zabelin, A.I.; Shmelev, V.E.

Radiolysis of the coolant proceeds at a higher rate in a boiling water reactor as compared to a water-moderated, water-cooled reactor. The radiolytic gases (hydrogen and oxygen) exiting the reactor together with steam can form a potentially explosive mixture. Special interest attaches to the results obtained under the codnitions of prolonged operation of the VK-50 reactor. Tests of various water-chemistry conditions which were performed in the experimental reactor showed their critical influence on the rate of progress of radiolytic processes. The entire period of operation of the reactor may be arbitrarily divided into three stages, each of which is characterizedmore » by its own peculiar conditions of water chemistry and range of thermal power. From stage to stage, there is a noticeable improvement in the coolant quality which to a limited extent is reflected in the exit of radiolytic gases with the steam. The concentration of radiolytic gases increases with decreased power and with an increased content of corrosion products and other contaminants in the coolant.« less

Experimental study on the heat transfer characteristics of a nuclear reactor containment wall cooled by gravitationally falling water

NASA Astrophysics Data System (ADS)

Pasek, Ari D.; Umar, Efrison; Suwono, Aryadi; Manalu, Reinhard E. E.

2012-06-01

Gravitationally falling water cooling is one of mechanism utilized by a modern nuclear Pressurized Water Reactor (PWR) for its Passive Containment Cooling System (PCCS). Since the cooling is closely related to the safety, water film cooling characteristics of the PCCS should be studied. This paper deals with the experimental study of laminar water film cooling on the containment model wall. The influences of water mass flow rate and wall heat rate on the heat transfer characteristic were studied. This research was started with design and assembly of a containment model equipped with the water cooling system, and calibration of all measurement devices. The containment model is a scaled down model of AP 1000 reactor. Below the containment steam is generated using electrical heaters. The steam heated the containment wall, and then the temperatures of the wall in several positions were measure transiently using thermocouples and data acquisition. The containment was then cooled by falling water sprayed from the top of the containment. The experiments were done for various wall heat rate and cooling water flow rate. The objective of the research is to find the temperature profile along the wall before and after the water cooling applied, prediction of the water film characteristic such as means velocity, thickness and their influence to the heat transfer coefficient. The result of the experiments shows that the wall temperatures significantly drop after being sprayed with water. The thickness of water film increases with increasing water flow rate and remained constant with increasing wall heat rate. The heat transfer coefficient decreases as film mass flow rate increase due to the increases of the film thickness which causes the increasing of the thermal resistance. The heat transfer coefficient increases slightly as the wall heat rate increases. The experimental results were then compared with previous theoretical studied.

Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

NASA Astrophysics Data System (ADS)

Cheng, Xiaoman; Ma, Xuebin; Jiang, Kecheng; Chen, Lei; Huang, Kai; Liu, Songlin

2015-09-01

The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

Catalytic reactor

DOEpatents

Aaron, Timothy Mark [East Amherst, NY; Shah, Minish Mahendra [East Amherst, NY; Jibb, Richard John [Amherst, NY

2009-03-10

A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

Pressure suppression system

DOEpatents

Gluntz, D.M.

1994-10-04

A pressure suppression system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and an enclosed gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The GDCS pool includes a plenum for receiving through an inlet the non-condensable gas carried with steam from the drywell following a loss-of-coolant accident (LOCA). A condenser is disposed in the GDCS plenum for condensing the steam channeled therein and to trap the non-condensable gas therein. A method of operation includes draining the GDCS pool following the LOCA and channeling steam released into the drywell following the LOCA into the GDCS plenum for cooling along with the non-condensable gas carried therewith for trapping the gas therein. 3 figs.

Pressure suppression system

DOEpatents

Gluntz, Douglas M.

1994-01-01

A pressure suppression system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and an enclosed gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The GDCS pool includes a plenum for receiving through an inlet the non-condensable gas carried with steam from the drywell following a loss-of-coolant accident (LOCA). A condenser is disposed in the GDCS plenum for condensing the steam channeled therein and to trap the non-condensable gas therein. A method of operation includes draining the GDCS pool following the LOCA and channeling steam released into the drywell following the LOCA into the GDCS plenum for cooling along with the non-condensable gas carried therewith for trapping the gas therein.

Passive containment cooling system

DOEpatents

Billig, P.F.; Cooke, F.E.; Fitch, J.R.

1994-01-25

A passive containment cooling system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel and is vented to the drywell. An isolation pool is disposed above the GDCS pool and includes an isolation condenser therein. The condenser has an inlet line disposed in flow communication with the drywell for receiving the non-condensable gas along with any steam released therein following a loss-of-coolant accident (LOCA). The condenser also has an outlet line disposed in flow communication with the drywell for returning to the drywell both liquid condensate produced upon cooling of the steam and the non-condensable gas for reducing pressure within the containment vessel following the LOCA. 1 figure.

Passive containment cooling system

DOEpatents

Billig, Paul F.; Cooke, Franklin E.; Fitch, James R.

1994-01-01

A passive containment cooling system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel and is vented to the drywell. An isolation pool is disposed above the GDCS pool and includes an isolation condenser therein. The condenser has an inlet line disposed in flow communication with the drywell for receiving the non-condensable gas along with any steam released therein following a loss-of-coolant accident (LOCA). The condenser also has an outlet line disposed in flow communication with the drywell for returning to the drywell both liquid condensate produced upon cooling of the steam and the non-condensable gas for reducing pressure within the containment vessel following the LOCA.

78 FR 8195 - Biweekly Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-05

... of the bases for the contention and a concise statement of the alleged facts or expert opinion which..., ``Allowable Value for Primary Containment and Drywell Isolation Instrumentation,'' Function 3.c, ``Reactor Core Isolation Cooling (RCIC) Steam Supply Line Pressure--Low.'' This TS allowable value will be...

Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility

NASA Astrophysics Data System (ADS)

Narcisi, V.; Giannetti, F.; Del Nevo, A.; Tarantino, M.; Caruso, G.

2017-11-01

In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermo-fluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation.

Westinghouse Small Modular Reactor balance of plant and supporting systems design

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 operationmore » 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)« less

Passive containment cooling system with drywell pressure regulation for boiling water reactor

DOEpatents

Hill, P.R.

1994-12-27

A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

Demonstrating Hybrid Heat Transport and Energy Conversion System Performance Characterization Using Intelligent Control Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ostrum, Lee; Manic, Milos

The debate continues on the magnitude and validity of climate change caused by human activities. However, there is no debate about the need to make buildings, modes of transportation, factories, and homes as energy efficient as possible. Given that climate change could occur with the wasteful use of fossil fuel and the fact that fossil energy costs could and will swing wildly, it is imperative that every effort be made to utilize energy sources to their fullest. Hybrid energy systems (HES) are two or more separate energy producers used together to produce energy commodities. The HES this report focuses onmore » is the use of nuclear reactor waste heat as a source of further energy utilization. Nuclear reactors use a fluid to cool the core and produce the steam needed for the production of electricity. Traditionally this steam, or coolant, is used to convert the energy then cooled elsewhere. The heat is released into the environment without being used further. By adding technologies to nuclear reactors to use the wasted heat, a system can be developed to make more than just electricity and allow for loading following capabilities.« less

Modeling Chilled-Water Storage System Components for Coupling to a Small Modular Reactor in a Nuclear Hybrid Energy System

NASA Astrophysics Data System (ADS)

Misenheimer, Corey Thomas

The intermittency of wind and solar power puts strain on electric grids, often forcing carbonbased and nuclear sources of energy to operate in a load-follow mode. Operating nuclear reactors in a load-follow fashion is undesirable due to the associated thermal and mechanical stresses placed on the fuel and other reactor components. Various Thermal Energy Storage (TES) elements and ancillary energy applications can be coupled to nuclear (or renewable) power sources to help absorb grid instabilities caused by daily electric demand changes and renewable intermittency, thereby forming the basis of a candidate Nuclear Hybrid Energy System (NHES). During the warmer months of the year in many parts of the country, facility air-conditioning loads are significant contributors to the increase in the daily peak electric demand. Previous research demonstrated that a stratified chilled-water storage tank can displace peak cooling loads to off-peak hours. Based on these findings, the objective of this work is to evaluate the prospect of using a stratified chilled-water storage tank as a potential TES reservoir for a nuclear reactor in a NHES. This is accomplished by developing time-dependent models of chilled-water system components, including absorption chillers, cooling towers, a storage tank, and facility cooling loads appropriate for a large office space or college campus, as a callable FORTRAN subroutine. The resulting TES model is coupled to a high-fidelity mPower-sized Small Modular Reactor (SMR) Simulator, with the goal of utilizing excess reactor capacity to operate several sizable chillers in order to keep reactor power constant. Chilled-water production via single effect, lithium bromide (LiBr) absorption chillers is primarily examined in this study, although the use of electric chillers is briefly explored. Absorption chillers use hot water or low-pressure steam to drive an absorption-refrigeration cycle. The mathematical framework for a high-fidelity dynamic absorption chiller model is presented. The transient FORTRAN model is grounded on time-dependent mass, species, and energy conservation equations. Due to the vast computational costs of the high-fidelity model, a low-fidelity absorption chiller model is formulated and calibrated to mimic the behavior of the high-fidelity model. Stratified chilled-water storage tank performance is characterized using Computational Fluid Dynamics (CFD). The geometry employed in the CFD model represents a 5-million-gallon storage tank currently in use at a North Carolina college campus. Simulation results reveal the laminar numerical model most closely aligns with actual tank charging and discharging data. A subsequent parametric study corroborates storage tank behavior documented throughout literature and industry. Two absorption chiller configurations are considered. The first involves bypassing lowpressure steam from the low-pressure turbine to absorption chillers during periods of excess reactor capacity in order to keep reactor power constant. Simulation results show steam conditions downstream of the turbine control valves are a strong function of turbine load, and absorption chiller performance is hindered by reduced turbine impulse pressures at reduced turbine demands. A more suitable configuration entails integrating the absorption chillers into a flash vessel system that is thermally coupled to a sensible heat storage system. The sensible heat storage system is able to maintain reactor thermal output constant at 100% and match turbine output with several different electric demand profiles. High-pressure condensate in the sensible heat storage system is dropped across a let-down orifice and flashed in an ideal separator. Generated steam is sent to a bank of absorption chillers. Simulation results show enough steam is available during periods of reduced turbine demand to power four large absorption chillers to charge a 5-million-gallon stratified chilled-water storage tank, which is used to offset cooling loads in an adjacent facility. The coupled TES systems operating in conjunction with an SMR comprise the foundation of a tightly coupled NHES.

A STRONGLY COUPLED REACTOR CORE ISOLATION COOLING SYSTEM MODEL FOR EXTENDED STATION BLACK-OUT ANALYSES

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2015-03-01

The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoidmore » overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety system components such as the safety relief valve (SRV), the RCIC system, the wet well, and the dry well. The results show reasonable system behaviors while exhibiting rich dynamics such as variable flow rates through RCIC turbine and pump during the SBO transient. The model has the potential to resolve the Fukushima RCIC mystery after adding the off-design two-phase turbine operation model and other additional improvements.« less

Multi-phase model development to assess RCIC system capabilities under severe accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kirkland, Karen Vierow; Ross, Kyle; Beeny, Bradley

The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that themore » system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.« less

Passive decay heat removal system for water-cooled nuclear reactors

DOEpatents

Forsberg, Charles W.

1991-01-01

A passive decay-heat removal system for a water-cooled nuclear reactor employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated box located inside the reactor vessel, below its normal water level, in communication with a condenser located outside of containment and exposed to the atmosphere. The heat transfer loop is located such that the evaporator is in a position where, when the water level drops in the reactor, it will become exposed to steam. Vapor produced in the evaporator passes upward to the condenser above the normal water level. In operation, condensation in the condenser removes heat from the system, and the condensed liquid is returned to the evaporator. The system is disposed such that during normal reactor operations where the water level is at its usual position, very little heat will be removed from the system, but during emergency, low water level conditions, substantial amounts of decay heat will be removed.

Corrosion Mechanisms in Steam Power Generation

DTIC Science & Technology

1960-01-01

Report #5427, Dec. 30, 1959 3. Hawkins, Agnew and Solberg, Trans. ASME 66:291 (1944) 4. M. C. Bloom and M. Krulfeld, J. Electrochem,1 Soc. 104: 264 ,(1957...Tackett, P. E. Brown and R.’T. Esper WAPD -LSR(C)-134 (October 14, 1955) 8. D. L. Douglas, Corrosion and Wear Handbook for Water-cooled Reactors

An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shannon M. Bragg-Sitton; Richard D. Boardman; Robert S. Cherry

2014-03-01

Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feedmore » a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hale, Richard Edward; Fugate, David L.; Cetiner, Sacit M.

2015-05-01

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

Investigation of the possibility of using residual heat reactor energy

NASA Astrophysics Data System (ADS)

Aminov, R. Z.; Yurin, V. E.; Bessonov, V. N.

2017-11-01

The largest contribution to the probable frequency of core damage is blackout events. The main component of the heat capacity at each reactor within a few minutes following a blackout is the heat resulting from the braking of beta-particles and the transfer of gamma-ray energy by the fission fragments and their decay products, which is known as the residual heat. The power of the residual heat changes gradually over a long period of time and for a VVER-1000 reactor is about 15-20 MW of thermal power over 72 hours. Current cooldown systems increase the cost of the basic nuclear power plants (NPP) funds without changing the amount of electricity generated. Such systems remain on standby, accelerating the aging of the equipment and accordingly reducing its reliability. The probability of system failure increases with the duration of idle time. Furthermore, the reactor residual heat energy is not used. A proposed system for cooling nuclear power plants involves the use of residual thermal power to supply the station’s own needs in emergency situations accompanied by a complete blackout. The thermal power of residual heat can be converted to electrical energy through an additional low power steam turbine. In normal mode, the additional steam turbine generates electricity, which makes it possible to ensure spare NPP and a return on the investment in the reservation system. In this work, experimental data obtained from a Balakovo NPP was analyzed to determine the admissibility of cooldown of the reactors through the 2nd circuit over a long time period, while maintaining high-level parameters for the steam generated by the steam generators.

NEUTRONIC REACTOR SYSTEM

DOEpatents

Daniels, F.

1957-10-15

Gas-cooled solid-moderator type reactors wherein the fissionable fuel and moderator materials are each in the form of solid pebbles, or discrete particles, and are substantially homogeneously mixed in the proper proportion and placed within the core of the reactor are described. The shape of these discrete particles must be such that voids are present between them when mixed together. Helium enters the bottom of the core and passes through the voids between the fuel and moderator particles to absorb the heat generated by the chain reaction. The hot helium gas is drawn off the top of the core and may be passed through a heat exchanger to produce steam.

Apparatus and methods of reheating gas turbine cooling steam and high pressure steam turbine exhaust in a combined cycle power generating system

DOEpatents

Tomlinson, Leroy Omar; Smith, Raub Warfield

2002-01-01

In a combined cycle system having a multi-pressure heat recovery steam generator, a gas turbine and steam turbine, steam for cooling gas turbine components is supplied from the intermediate pressure section of the heat recovery steam generator supplemented by a portion of the steam exhausting from the HP section of the steam turbine, steam from the gas turbine cooling cycle and the exhaust from the HP section of the steam turbine are combined for flow through a reheat section of the HRSG. The reheated steam is supplied to the IP section inlet of the steam turbine. Thus, where gas turbine cooling steam temperature is lower than optimum, a net improvement in performance is achieved by flowing the cooling steam exhausting from the gas turbine and the exhaust steam from the high pressure section of the steam turbine in series through the reheater of the HRSG for applying steam at optimum temperature to the IP section of the steam turbine.

Rotating diffuser for pressure recovery in a steam cooling circuit of a gas turbine

DOEpatents

Eldrid, Sacheverel Q.; Salamah, Samir A.; DeStefano, Thomas Daniel

2002-01-01

The buckets of a gas turbine are steam-cooled via a bore tube assembly having concentric supply and spent cooling steam return passages rotating with the rotor. A diffuser is provided in the return passage to reduce the pressure drop. In a combined cycle system, the spent return cooling steam with reduced pressure drop is combined with reheat steam from a heat recovery steam generator for flow to the intermediate pressure turbine. The exhaust steam from the high pressure turbine of the combined cycle unit supplies cooling steam to the supply conduit of the gas turbine.

Exploratory study of several advanced nuclear-MHD power plant systems.

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.; Rosa, R. J.; Yang, Y. Y.

1973-01-01

In order for efficient multimegawatt closed cycle nuclear-MHD systems to become practical, long-life gas cooled reactors with exit temperatures of about 2500 K or higher must be developed. Four types of nuclear reactors which have the potential of achieving this goal are the NERVA-type solid core reactor, the colloid core (rotating fluidized bed) reactor, the 'light bulb' gas core reactor, and the 'coaxial flow' gas core reactor. Research programs aimed at developing these reactors have progressed rapidly in recent years so that prototype power reactors could be operating by 1980. Three types of power plant systems which use these reactors have been analyzed to determine the operating characteristics, critical parameters and performance of these power plants. Overall thermal efficiencies as high as 80% are projected, using an MHD turbine-compressor cycle with steam bottoming, and slightly lower efficiencies are projected for an MHD motor-compressor cycle.

Posttest analysis of MIST Test 3109AA using TRAC-PF1/MOD1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steiner, J.L.; Siebe, D.A.; Boyack, B.E.

This document discusses a posttest calculation and analysis of Multi-loop Integral System Test (MIST) 3109AA as the nominal test for the MIST program. It is a test of a small-break loss-of-coolant accident (SBLOCA) with a scaled 10-cm{sup 2} break in the B1 cold leg. The test exhibited the major post-SBLOCA phenomena, as expected, including depressurization to saturation, intermittent and interrupted loop flow, boiler-condenser mode cooling, refill, and postrefill cooldown. Full high-pressure injection and auxiliary feedwater were available, reactor coolant pumps were not available, and reactor-vessel vent valves and guard heaters were automatically controlled. Constant level control in the steam-generator secondariesmore » was used after steam-generator secondary refill and symmetric steam-generator pressure control was used. We performed the calculation using TRAC-PF1/MODI. Agreement between test data and the calculation was generally reasonable. All major trends and phenomena were correctly predicted. It is believed that the correct conclusions about trends and phenomena will be reached if the code is used in similar applications.« less

Posttest analysis of MIST Test 3109AA using TRAC-PF1/MOD1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steiner, J.L.; Siebe, D.A.; Boyack, B.E.

This document discusses a posttest calculation and analysis of Multi-loop Integral System Test (MIST) 3109AA as the nominal test for the MIST program. It is a test of a small-break loss-of-coolant accident (SBLOCA) with a scaled 10-cm[sup 2] break in the B1 cold leg. The test exhibited the major post-SBLOCA phenomena, as expected, including depressurization to saturation, intermittent and interrupted loop flow, boiler-condenser mode cooling, refill, and postrefill cooldown. Full high-pressure injection and auxiliary feedwater were available, reactor coolant pumps were not available, and reactor-vessel vent valves and guard heaters were automatically controlled. Constant level control in the steam-generator secondariesmore » was used after steam-generator secondary refill and symmetric steam-generator pressure control was used. We performed the calculation using TRAC-PF1/MODI. Agreement between test data and the calculation was generally reasonable. All major trends and phenomena were correctly predicted. It is believed that the correct conclusions about trends and phenomena will be reached if the code is used in similar applications.« less

Optimization of 200 MWth and 250 MWt Ship Based Small Long Life NPP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fitriyani, Dian; Su'ud, Zaki

2010-06-22

Design optimization of ship-based 200 MWth and 250 MWt nuclear power reactors have been performed. The neutronic and thermo-hydraulic programs of the three-dimensional X-Y-Z geometry have been developed for the analysis of ship-based nuclear power plant. Quasi-static approach is adopted to treat seawater effect. The reactor are loop type lead bismuth cooled fast reactor with nitride fuel and with relatively large coolant pipe above reactor core, the heat from primary coolant system is directly transferred to watersteam loop through steam generators. Square core type are selected and optimized. As the optimization result, the core outlet temperature distribution is changing withmore » the elevation angle of the reactor system and the characteristics are discussed.« less

Passive Acoustic Leak Detection for Sodium Cooled Fast Reactors Using Hidden Markov Models

NASA Astrophysics Data System (ADS)

Marklund, A. Riber; Kishore, S.; Prakash, V.; Rajan, K. K.; Michel, F.

2016-06-01

Acoustic leak detection for steam generators of sodium fast reactors have been an active research topic since the early 1970s and several methods have been tested over the years. Inspired by its success in the field of automatic speech recognition, we here apply hidden Markov models (HMM) in combination with Gaussian mixture models (GMM) to the problem. To achieve this, we propose a new feature calculation scheme, based on the temporal evolution of the power spectral density (PSD) of the signal. Using acoustic signals recorded during steam/water injection experiments done at the Indira Gandhi Centre for Atomic Research (IGCAR), the proposed method is tested. We perform parametric studies on the HMM+GMM model size and demonstrate that the proposed method a) performs well without a priori knowledge of injection noise, b) can incorporate several noise models and c) has an output distribution that simplifies false alarm rate control.

Hydrogen production from coal using a nuclear heat source

NASA Technical Reports Server (NTRS)

Quade, R. N.

1976-01-01

A strong candidate for hydrogen production in the intermediate time frame of 1985 to 1995 is a coal-based process using a high-temperature gas-cooled reactor (HTGR) as a heat source. Expected process efficiencies in the range of 60 to 70% are considerably higher than all other hydrogen production processes except steam reforming of a natural gas. The process involves the preparation of a coal liquid, hydrogasification of that liquid, and steam reforming of the resulting gaseous or light liquid product. A study showing process efficiency and cost of hydrogen vs nuclear reactor core outlet temperature has been completed, and shows diminishing returns at process temperatures above about 1500 F. A possible scenario combining the relatively abundant and low-cost Western coal deposits with the Gulf Coast hydrogen users is presented which provides high-energy density transportation utilizing coal liquids and uranium.

Locating hot and cold-legs in a nuclear powered steam generation system

DOEpatents

Ekeroth, D.E.; Corletti, M.M.

1993-11-16

A nuclear reactor steam generator includes a reactor vessel for heating water and a steam generator with a pump casing at the lowest point on the steam generator. A cold-leg pipe extends horizontally between the steam generator and the reactor vessel to return water from the steam generator to the reactor vessel. The bottom of the cold-leg pipe is at a first height above the bottom of the reactor vessel. A hot-leg pipe with one end connected to the steam generator and a second end connected to the reactor vessel has a first pipe region extending downwardly from the steam generator to a location between the steam generator and the reactor vessel at which a bottom of the hot-leg pipe is at a second height above the bottom of the reactor vessel. A second region extends from that location in a horizontal direction at the second height to the point at which the hot-leg pipe connects to the reactor vessel. A pump is attached to the casing at a location below the first and second heights and returns water from the steam generator to the reactor vessel over the cold-leg. The first height is greater than the second height and the bottom of the steam generator is at a height above the bottom of the reactor vessel that is greater than the first and second heights. A residual heat recovery pump is below the hot-leg and has an inlet line from the hot-leg that slopes down continuously to the pump inlet. 2 figures.

Locating hot and cold-legs in a nuclear powered steam generation system

DOEpatents

Ekeroth, Douglas E.; Corletti, Michael M.

1993-01-01

A nuclear reactor steam generator includes a reactor vessel for heating water and a steam generator with a pump casing at the lowest point on the steam generator. A cold-leg pipe extends horizontally between the steam generator and the reactor vessel to return water from the steam generator to the reactor vessel. The bottom of the cold-leg pipe is at a first height above the bottom of the reactor vessel. A hot-leg pipe with one end connected to the steam generator and a second end connected to the reactor vessel has a first pipe region extending downwardly from the steam generator to a location between the steam generator and the reactor vessel at which a bottom of the hot-leg pipe is at a second height above the bottom of the reactor vessel. A second region extends from that location in a horizontal direction at the second height to the point at which the hot-leg pipe connects to the reactor vessel. A pump is attached to the casing at a location below the first and second heights and returns water from the steam generator to the reactor vessel over the cold-leg. The first height is greater than the second height and the bottom of the steam generator is at a height above the bottom of the reactor vessel that is greater than the first and second heights. A residual heat recovery pump is below the hot-leg and has an inlet line from the hot-leg that slopes down continuously to the pump inlet.

Pressure suppression containment system

DOEpatents

Gluntz, Douglas M.; Townsend, Harold E.

1994-03-15

A pressure suppression containment system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The wetwell pool includes a plenum for receiving the non-condensable gas carried with steam from the drywell following a loss-of coolant-accident (LOCA). The wetwell plenum is vented to a plenum above the GDCS pool following the LOCA for suppressing pressure rise within the containment vessel. A method of operation includes channeling steam released into the drywell following the LOCA into the wetwell pool for cooling along with the non-condensable gas carried therewith. The GDCS pool is then drained by gravity, and the wetwell plenum is vented into the GDCS plenum for channeling the non-condensable gas thereto.

Pressure suppression containment system

DOEpatents

Gluntz, D.M.; Townsend, H.E.

1994-03-15

A pressure suppression containment system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The wetwell pool includes a plenum for receiving the non-condensable gas carried with steam from the drywell following a loss-of-coolant-accident (LOCA). The wetwell plenum is vented to a plenum above the GDCS pool following the LOCA for suppressing pressure rise within the containment vessel. A method of operation includes channeling steam released into the drywell following the LOCA into the wetwell pool for cooling along with the non-condensable gas carried therewith. The GDCS pool is then drained by gravity, and the wetwell plenum is vented into the GDCS plenum for channeling the non-condensable gas thereto. 6 figures.

Apparatus and methods for supplying auxiliary steam in a combined cycle system

DOEpatents

Gorman, William G.; Carberg, William George; Jones, Charles Michael

2002-01-01

To provide auxiliary steam, a low pressure valve is opened in a combined cycle system to divert low pressure steam from the heat recovery steam generator to a header for supplying steam to a second combined cycle's steam turbine seals, sparging devices and cooling steam for the steam turbine if the steam turbine and gas turbine lie on a common shaft with the generator. Cooling steam is supplied the gas turbine in the combined cycle system from the high pressure steam turbine. Spent gas turbine cooling steam may augment the low pressure steam supplied to the header by opening a high pressure valve whereby high and low pressure steam flows are combined. An attemperator is used to reduce the temperature of the combined steam in response to auxiliary steam flows above a predetermined flow and a steam header temperature above a predetermined temperature. The auxiliary steam may be used to start additional combined cycle units or to provide a host unit with steam turbine cooling and sealing steam during full-speed no-load operation after a load rejection.

The IRIS Spool-Type Reactor Coolant Pump

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kujawski, J.M.; Kitch, D.M.; Conway, L.E.

2002-07-01

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safetymore » to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called 'spool' pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper. (authors)« less

Amoeba behavior of UO/sub 2/ coated particle fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagner-Loeffler, M.

1977-09-01

The data extracted from numerous irradiation tests were used to derive amoeba endurance parameters for UO/sub 2/. The data do not yet allow an unambiguous definition of the controlling mechanism, which may be due to either gaseous or solid-state diffusion processes. Adequate data on the amoeba effect are available for design of a steam-raising high-temperature gas-cooled reactor using UO/sub 2/ fuel.

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... schedule for providing a reanalysis or taking other action as may be needed to show compliance with § 50.46... schedule for providing a reanalysis or taking other action as may be needed to show compliance with § 50.46... generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the...

Debris trap in a turbine cooling system

DOEpatents

Wilson, Ian David

2002-01-01

In a turbine having a rotor and a plurality of stages, each stage comprising a row of buckets mounted on the rotor for rotation therewith; and wherein the buckets of at least one of the stages are cooled by steam, the improvement comprising at least one axially extending cooling steam supply conduit communicating with an at least partially annular steam supply manifold; one or more axially extending cooling steam feed tubes connected to the manifold at a location radially outwardly of the cooling steam supply conduit, the feed tubes arranged to supply cooling steam to the buckets of at least one of the plurality of stages; the manifold extending radially beyond the feed tubes to thereby create a debris trap region for collecting debris under centrifugal loading caused by rotation of the rotor.

SUPERHEATING IN A BOILING WATER REACTOR

DOEpatents

Treshow, M.

1960-05-31

A boiling-water reactor is described in which the steam developed in the reactor is superheated in the reactor. This is accomplished by providing means for separating the steam from the water and passing the steam over a surface of the fissionable material which is not in contact with the water. Specifically water is boiled on the outside of tubular fuel elements and the steam is superheated on the inside of the fuel elements.

High efficiency Brayton cycles using LNG

DOEpatents

Morrow, Charles W [Albuquerque, NM

2006-04-18

A modified, closed-loop Brayton cycle power conversion system that uses liquefied natural gas as the cold heat sink media. When combined with a helium gas cooled nuclear reactor, achievable efficiency can approach 68 76% (as compared to 35% for conventional steam cycle power cooled by air or water). A superheater heat exchanger can be used to exchange heat from a side-stream of hot helium gas split-off from the primary helium coolant loop to post-heat vaporized natural gas exiting from low and high-pressure coolers. The superheater raises the exit temperature of the natural gas to close to room temperature, which makes the gas more attractive to sell on the open market. An additional benefit is significantly reduced costs of a LNG revaporization plant, since the nuclear reactor provides the heat for vaporization instead of burning a portion of the LNG to provide the heat.

Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

DOEpatents

McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

1994-05-03

The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

DOEpatents

McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

1994-01-01

The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

5. STEAM PLANT COOLING TOWER LOCATED WEST OF STEAM PLANT ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

5. STEAM PLANT COOLING TOWER LOCATED WEST OF STEAM PLANT BUILDING, FROM SOUTH. SHOWS CURRENT LEVEL OF DISREPAIR. December 4, 1990 - Crosscut Steam Plant, North side Salt River near Mill Avenue & Washington Street, Tempe, Maricopa County, AZ

Regulatory Concerns on the In-Containment Water Storage System of the Korean Next Generation Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahn, Hyung-Joon; Lee, Jae-Hun; Bang, Young-Seok

2002-07-15

The in-containment water storage system (IWSS) is a newly adopted system in the design of the Korean Next Generation Reactor (KNGR). It consists of the in-containment refueling water storage tank, holdup volume tank, and cavity flooding system (CFS). The IWSS has the function of steam condensation and heat sink for the steam release from the pressurizer and provides cooling water to the safety injection system and containment spray system in an accident condition and to the CFS in a severe accident condition. With the progress of the KNGR design, the Korea Institute of Nuclear Safety has been developing Safety andmore » Regulatory Requirements and Guidances for safety review of the KNGR. In this paper, regarding the IWSS of the KNGR, the major contents of the General Safety Criteria, Specific Safety Requirements, Safety Regulatory Guides, and Safety Review Procedures were introduced, and the safety review items that have to be reviewed in-depth from the regulatory viewpoint were also identified.« less

Passive acoustic leak detection for sodium cooled fast reactors using hidden Markov models

DOE Office of Scientific and Technical Information (OSTI.GOV)

Riber Marklund, A.; Kishore, S.; Prakash, V.

2015-07-01

Acoustic leak detection for steam generators of sodium fast reactors have been an active research topic since the early 1970's and several methods have been tested over the years. Inspired by its success in the field of automatic speech recognition, we here apply hidden Markov models (HMM) in combination with Gaussian mixture models (GMM) to the problem. To achieve this, we propose a new feature calculation scheme, based on the temporal evolution of the power spectral density (PSD) of the signal. Using acoustic signals recorded during steam/water injection experiments done at the Indira Gandhi Centre for Atomic Research (IGCAR), themore » proposed method is tested. We perform parametric studies on the HMM+GMM model size and demonstrate that the proposed method a) performs well without a priori knowledge of injection noise, b) can incorporate several noise models and c) has an output distribution that simplifies false alarm rate control. (authors)« less

CHARACTERISTIC QUALITIES OF SOME ATOMIC POWER STATIONS (in Hungarian)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ligeti, G.

1962-04-01

Mostly as the result of economic factors, the current rate of construction of public atomic power stations has slowed down. The use of atomic energy is considered economical only in a few special cases, such as ship propulsion or supplying power to remote regions. For this reason, many reactors were designed especially for the construction of such midget'' power stations, operating at power levels ranging from 10 to 70 Mw. Technical details are given of such already-built or proposed systems, including the following: pressurized- water reactors such as the Babcock and Wilcox 60-Mw reactor, using 2.4% U/sup 235/ fuel; themore » Humphrey-Glasow Company's 20 Mw reactor; the gascooled system of the de Havilland Company; the organicmoderated reactor of the English Electric Company; the organic-moderated system of the Hawker-Siddeley Nuclear Power Company; the boiling-water reactor of the Mitchell Engineering Company and the steam-cooled, heavy-water reactor of the Rolls-Royce & Vickers Company. (TTT)« less

STEAM GENERATOR FOR NUCLEAR REACTOR

DOEpatents

Kinyon, B.W.; Whitman, G.D.

1963-07-16

The steam generator described for use in reactor powergenerating systems employs a series of concentric tubes providing annular passage of steam and water and includes a unique arrangement for separating the steam from the water. (AEC)

Closed circuit steam cooled turbine shroud and method for steam cooling turbine shroud

DOEpatents

Burdgick, Steven Sebastian; Sexton, Brendan Francis; Kellock, Iain Robertson

2002-01-01

A turbine shroud cooling cavity is partitioned to define a plurality of cooling chambers for sequentially receiving cooling steam and impingement cooling of the radially inner wall of the shoud. An impingement baffle is provided in each cooling chamber for receiving the cooling media from a cooling media inlet in the case of the first chamber or from the immediately upstream chamber in the case of the second through fourth chambers and includes a plurality of impingement holes for effecting the impingement cooling of the shroud inner wall.

Bed Agglomeration During the Steam Gasification of a High Lignin Corn Stover Simultaneous Saccharification and Fermentation (SSF) Digester Residue

DOE Office of Scientific and Technical Information (OSTI.GOV)

Howe, Daniel T.; Taasevigen, Danny J.; Gerber, Mark A.

This research investigates the bed agglomeration phenomena during the steam gasification of a high lignin residue produced from the simultaneous saccharification and fermentation (SSF) of corn stover in a bubbling fluidized bed. The studies were conducted at 895°C using alumina as bed material. Biomass was fed at 1.5 kg/hr, while steam was fed to give a velocity equal to 2.5 times the minimum fluidization velocity, with a steam/carbon ratio of 0.9. The pelletized feedstock was co-fed with a cooling nitrogen stream to mitigate feed line plugging issues. Tar production was high at 50.3 g/Nm3, and the fraction of C10+ compoundsmore » was greater than that seen in the gasification of traditional lignocellulosic feedstocks. Carbon closures over 94 % were achieved for all experiments. Bed agglomeration was found to be problematic, indicated by pressure drop increases observed below the bed and upstream of the feed line. Two size categories of solids were recovered from the reactor, +60 mesh and -60 mesh. After a 2.75-hour experiment, 61.7 wt % was recovered as -60 mesh particles and 38.2 wt% of the recovered reactor solids were +60 mesh. A sizeable percentage, 31.8 wt%, was +20 mesh. The -60 mesh particles were mainly formed by the initial bed material (Al2O3). Almost 50 wt. % of the + 20 mesh particles was found to be formed by organics. The unreacted carbon remaining in the reactor resulted in a low conversion rate to product gas. ICP-AES, SEM, SEM-EDS, and XRD confirmed that the large agglomerates (+ 20 mesh) were not encapsulated bed material but rather un-gasified feedstock pellets with sand particles attached to it.« less

The Fukushima Nuclear Disaster and the U.S. Customs and Border Protection Response

NASA Astrophysics Data System (ADS)

McCormick, Kathy

2013-10-01

On 3/11/11, the reactors at the Fukushima Nuclear Plant in Japan were damaged by a magnitude 9.0 earthquake. Of the six reactors at the site, three were in operation prior to the event, and were automatically shut-down during the earthquake. Emergency cooling systems came online and were subsequently destroyed by a tsunami generated by the earthquake. For the operating reactors, all the reactor cores were exposed, resulting in overheating and the release of steam and hydrogen gas to the containment vessels, several of which subsequently exploded, releasing radioactivity into the atmosphere. The cores of the operating reactors melted down, and radioactive water was released to the ocean in cooling efforts. The primary radiation concerns in the United States from the disaster were radioactive plumes driven by westerly winds and contaminated commercial products and travelers. In the United States, one of the primary governmental organizations to respond to the disaster was U.S. Customs and Border Protection (CBP), which has responsibility to oversee the safety and security of cargo and travelers entering the United States. This talk will describe the various types of radioactive commodities and events encountered by CBP in the U.S. from the Fukushima disaster. Thanks to the CBP Teleforensics Center for their assistance with this presentation.

Cooling system for a bearing of a turbine rotor

DOEpatents

Schmidt, Mark Christopher

2002-01-01

In a gas turbine, a bore tube assembly radially inwardly of an aft bearing conveys cooling steam to the buckets of the turbine and returns the cooling steam to a return. To cool the bearing and thermally insulate the bearing from the cooling steam paths, a radiation shield is spaced from the bore tube assembly by a dead air gap. Additionally, an air passageway is provided between the radiation shield and the inner surface of an aft shaft forming part of the rotor. Air is supplied from an inlet for flow along the passage and radially outwardly through bores in the aft shaft disk to cool the bearing and insulate it from transfer of heat from the cooling steam.

Features of steam turbine cooling by the example of an SKR-100 turbine for supercritical steam parameters

NASA Astrophysics Data System (ADS)

Arkadyev, B. A.

2015-10-01

Basic principles of cooling of high-temperature steam turbines and constructive solutions used for development of the world's first cooled steam turbine SKR-100 (R-100-300) are described. Principal differences between the thermodynamic properties of cooling medium in the steam and gas turbines and the preference of making flow passes of cooled cylinders of steam turbines as reactive are shown. Some of its operation results and their conclusions are given. This turbine with a power of 100 MW, initial steam parameters approximately 30 MPa and 650°C, and back pressure 3 MPa was made by a Kharkov turbine plant in 1961 and ran successfully at a Kashira GRES (state district power plant) up to 1979, when it was taken out of use in a still fully operating condition. For comparison, some data on construction features and operation results of the super-high pressure cylinder of steam turbines of American Philo 6 (made by General Electric Co.) and Eddystone 1 (made by Westinghouse Co.) power generating units, which are close to the SKR-100 turbine by design initial steam parameters and the implementation time, are given. The high operational reliability and effectiveness of the cooling system that was used in the super-high pressure cylinder of the SKR-100 turbine of the power-generating unit, which were demonstrated in operation, confirms rightfulness and expediency of principles and constructive solutions laid at its development. As process steam temperatures are increased, the realization of the proposed approach to cooling of multistage turbines makes it possible to limit for large turbine parts the application of new, more expensive high-temperature materials, which are required for making steam boilers, and, in some cases, to do completely away with their utilization.

Steam exit flow design for aft cavities of an airfoil

DOEpatents

Storey, James Michael; Tesh, Stephen William

2002-01-01

Turbine stator vane segments have inner and outer walls with vanes extending therebetween. The inner and outer walls have impingement plates. Steam flowing into the outer wall passes through the impingement plate for impingement cooling of the outer wall surface. The spent impingement steam flows into cavities of the vane having inserts for impingement cooling the walls of the vane. The steam passes into the inner wall and through the impingement plate for impingement cooling of the inner wall surface and for return through return cavities having inserts for impingement cooling of the vane surfaces. A skirt or flange structure is provided for shielding the steam cooling impingement holes adjacent the inner wall aerofoil fillet region of the nozzle from the steam flow exiting the aft nozzle cavities. Moreover, the gap between the flash rib boss and the cavity insert is controlled to minimize the flow of post impingement cooling media therebetween. This substantially confines outflow to that exiting via the return channels, thus furthermore minimizing flow in the vicinity of the aerofoil fillet region that may adversely affect impingement cooling thereof.

Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grabaskas, D.; Brunett, A.; Passerini, S.

Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. Themore » mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.« less

Carbonaceous material for production of hydrogen from low heating value fuel gases

DOEpatents

Koutsoukos, Elias P.

1989-01-01

A process for the catalytic production of hydrogen, from a wide variety of low heating value fuel gases containing carbon monoxide, comprises circulating a carbonaceous material between two reactors--a carbon deposition reactor and a steaming reactor. In the carbon deposition reactor, carbon monoxide is removed from a fuel gas and is deposited on the carbonaceous material as an active carbon. In the steaming reactor, the reactive carbon reacts with steam to give hydrogen and carbon dioxide. The carbonaceous material contains a metal component comprising from about 75% to about 95% cobalt, from about 5% to about 15% iron, and up to about 10% chromium, and is effective in suppressing the production of methane in the steaming reactor.

CONTEXTUAL AERIAL VIEW OF "COLD" NORTH HALF OF MTR COMPLEX. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

CONTEXTUAL AERIAL VIEW OF "COLD" NORTH HALF OF MTR COMPLEX. CAMERA FACING EASTERLY. FOREGROUND CORNER CONTAINS OIL STORAGE TANKS. WATER TANKS AND WELL HOUSES ARE BEYOND THEM TO THE LEFT. LARGE LIGHT-COLORED BUILDING IN CENTER OF VIEW IS STEAM PLANT. DEMINERALIZER AND WATER STORAGE TANK ARE BEYOND. SIX-CELL COOLING TOWER AND ITS PUMP HOUSE ARE ABOVE IT IN VIEW. SERVICE BUILDINGS INCLUDING CANTEEN ARE ON NORTH SIDE OF ROAD. "EXCLUSION" AREA IS BEYOND ROAD. COMPARE LOCATION OF EXCLUSION-AREA GATE WITH PHOTO ID-33-G-202. INL NEGATIVE NO. 3608. Unknown Photographer, 10/30/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Steam Reformer With Fibrous Catalytic Combustor

NASA Technical Reports Server (NTRS)

Voecks, Gerald E.

1987-01-01

Proposed steam-reforming reactor derives heat from internal combustion on fibrous catalyst. Supplies of fuel and air to combustor controlled to meet demand for heat for steam-reforming reaction. Enables use of less expensive reactor-tube material by limiting temperature to value safe for material yet not so low as to reduce reactor efficiency.

Cooling circuit for steam and air-cooled turbine nozzle stage

DOEpatents

Itzel, Gary Michael; Yu, Yufeng

2002-01-01

The turbine vane segment includes inner and outer walls with a vane extending therebetween. The vane includes leading and trailing edge cavities and intermediate cavities. An impingement plate is spaced from the outer wall to impingement-cool the outer wall. Post-impingement cooling air flows through holes in the outer wall to form a thin air-cooling film along the outer wall. Cooling air is supplied an insert sleeve with openings in the leading edge cavity for impingement-cooling the leading edge. Holes through the leading edge afford thin-film cooling about the leading edge. Cooling air is provided the trailing edge cavity and passes through holes in the side walls of the vane for thin-film cooling of the trailing edge. Steam flows through a pair of intermediate cavities for impingement-cooling of the side walls. Post-impingement steam flows to the inner wall for impingement-cooling of the inner wall and returns the post-impingement cooling steam through inserts in other intermediate cavities for impingement-cooling the side walls of the vane.

STEAM STIRRED HOMOGENEOUS NUCLEAR REACTOR

DOEpatents

Busey, H.M.

1958-06-01

A homogeneous nuclear reactor utilizing a selfcirculating liquid fuel is described. The reactor vessel is in the form of a vertically disposed tubular member having the lower end closed by the tube walls and the upper end closed by a removal fianged assembly. A spherical reaction shell is located in the lower end of the vessel and spaced from the inside walls. The reaction shell is perforated on its lower surface and is provided with a bundle of small-diameter tubes extending vertically upward from its top central portion. The reactor vessel is surrounded in the region of the reaction shell by a neutron reflector. The liquid fuel, which may be a solution of enriched uranyl sulfate in ordinary or heavy water, is mainiained at a level within the reactor vessel of approximately the top of the tubes. The heat of the reaction which is created in the critical region within the spherical reaction shell forms steam bubbles which more upwardly through the tubes. The upward movement of these bubbles results in the forcing of the liquid fuel out of the top of these tubes, from where the fuel passes downwardly in the space between the tubes and the vessel wall where it is cooled by heat exchangers. The fuel then re-enters the critical region in the reaction shell through the perforations in the bottom. The upper portion of the reactor vessel is provided with baffles to prevent the liquid fuel from splashing into this region which is also provided with a recombiner apparatus for recombining the radiolytically dissociated moderator vapor and a control means.

Steam cooling system for a gas turbine

DOEpatents

Wilson, Ian David; Barb, Kevin Joseph; Li, Ming Cheng; Hyde, Susan Marie; Mashey, Thomas Charles; Wesorick, Ronald Richard; Glynn, Christopher Charles; Hemsworth, Martin C.

2002-01-01

The steam cooling circuit for a gas turbine includes a bore tube assembly supplying steam to circumferentially spaced radial tubes coupled to supply elbows for transitioning the radial steam flow in an axial direction along steam supply tubes adjacent the rim of the rotor. The supply tubes supply steam to circumferentially spaced manifold segments located on the aft side of the 1-2 spacer for supplying steam to the buckets of the first and second stages. Spent return steam from these buckets flows to a plurality of circumferentially spaced return manifold segments disposed on the forward face of the 1-2 spacer. Crossover tubes couple the steam supply from the steam supply manifold segments through the 1-2 spacer to the buckets of the first stage. Crossover tubes through the 1-2 spacer also return steam from the buckets of the second stage to the return manifold segments. Axially extending return tubes convey spent cooling steam from the return manifold segments to radial tubes via return elbows.

On the possibility of connecting a non-operating main circulation pump with three pumps in operation without preliminary coast-down of power-generating unit No. 5 in the Novovoronezh nuclear power plant

NASA Astrophysics Data System (ADS)

Vitkovskii, I. L.; Nikonov, S. P.; Ryasnyi, S. I.

2014-02-01

The subject of this paper is a transient caused by connection of a standby loop to three operating circulation pumps at the initial reactor heat rate equal to 70% of the rated value without preliminarily reducing it to 30% of the rated level as required by the safe operation regulations. Failure of the following normal operation systems is supposed: the first- and the second-type warning protection systems, all quick-acting reducing devices releasing steam into the auxiliary manifold, the electric heaters of the pressurizer, the pressurizer injection system, the primary cooling circuit fluid makeup/blow-through systems, and the blocking systems to shut down the main circulation pump after the level in the steam generator is exceeded. In addition, it is supposed that, under transient conditions, the valves of the turbine regulation system will be in the position in which they were at the moment of the initial event until generation of the signal for positive closing of the turbine stop valves. The first signal to actuate the reactor emergency protection system (EPS) is skipped. The failure of all quick-acting reducing devices releasing steam into the atmosphere is assumed. In addition to equipment failure, at the moment when the main circulation pump is connected, the operator erroneously puts in a new setting to maintain the power allowable for four pumps in operation-in the calculations it was taken equal to 104% of the rated level at most considering the accuracy of evaluating and maintaining the reactor heat rate-and the working group of the reactor protection and control system (P&CS) starts moving upward. On reaching the set power level, the automatic reactor power regulator stops operating and the P&CS elements remain in the position in which they are at the moment. Compliance with the design safety criteria for the adopted scenario of the transient is demonstrated.

An numerical analysis of high-temperature helium reactor power plant for co-production of hydrogen and electricity

NASA Astrophysics Data System (ADS)

Dudek, M.; Podsadna, J.; Jaszczur, M.

2016-09-01

In the present work, the feasibility of using a high temperature gas cooled nuclear reactor (HTR) for electricity generation and hydrogen production are analysed. The HTR is combined with a steam and a gas turbine, as well as with the system for heat delivery for medium temperature hydrogen production. Industrial-scale hydrogen production using copper-chlorine (Cu-Cl) thermochemical cycle is considered and compared with high temperature electrolysis. Presented cycle shows a very promising route for continuous, efficient, large-scale and environmentally benign hydrogen production without CO2 emissions. The results show that the integration of a high temperature helium reactor, with a combined cycle for electric power generation and hydrogen production, may reach very high efficiency and could possibly lead to a significant decrease of hydrogen production costs.

Bore tube assembly for steam cooling a turbine rotor

DOEpatents

DeStefano, Thomas Daniel; Wilson, Ian David

2002-01-01

An axial bore tube assembly for a turbine is provided to supply cooling steam to hot gas components of the turbine wheels and return the spent cooling steam. A pair of inner and outer tubes define a steam supply passage concentric about an inner return passage. The forward ends of the tubes communicate with an end cap assembly having sets of peripheral holes communicating with first and second sets of radial tubes whereby cooling steam from the concentric passage is supplied through the end cap holes to radial tubes for cooling the buckets and return steam from the buckets is provided through the second set of radial tubes through a second set of openings of the end cap into the coaxial return passage. A radial-to-axial flow transitioning device, including anti-swirling vanes is provided in the end cap. A strut ring adjacent the aft end of the bore tube assembly permits axial and radial thermal expansion of the inner tube relative to the outer tube.

RELAP5-3D Modeling of Heat Transfer Components (Intermediate Heat Exchanger and Helical-Coil Steam Generator) for NGNP Application

DOE Office of Scientific and Technical Information (OSTI.GOV)

N. A. Anderson; P. Sabharwall

2014-01-01

The Next Generation Nuclear Plant project is aimed at the research and development of a helium-cooled high-temperature gas reactor that could generate both electricity and process heat for the production of hydrogen. The heat from the high-temperature primary loop must be transferred via an intermediate heat exchanger to a secondary loop. Using RELAP5-3D, a model was developed for two of the heat exchanger options a printed-circuit heat exchanger and a helical-coil steam generator. The RELAP5-3D models were used to simulate an exponential decrease in pressure over a 20 second period. The results of this loss of coolant analysis indicate thatmore » heat is initially transferred from the primary loop to the secondary loop, but after the decrease in pressure in the primary loop the heat is transferred from the secondary loop to the primary loop. A high-temperature gas reactor model should be developed and connected to the heat transfer component to simulate other transients.« less

Rotor bore and turbine rotor wheel/spacer heat exchange flow circuit

DOEpatents

Caruso, Philip M.; Eldrid, Sacheverel Quentin; Ladhani, Azad A.; DeMania, Alan Richard; Palmer, Gene David; Wilson, Ian David; Rathbun, Lisa Shirley; Akin, Robert Craig

2002-01-01

In a turbine having closed-circuit steam-cooling passages about the rim of the rotor during steady-state operation, compressor discharge air is supplied to the rotor bore for passage radially outwardly into the wheel space cavities between the wheels and spacers. Communicating slots and channels in the spacers and wheels at circumferentially spaced positions enable egress of the compressor discharge air into the hot gas flow path. At turbine startup, cooling air flows through the closed-circuit steam passages to cool the outer rim of the rotor while compressor discharge air pre-warms the wheels and spacers. At steady-state, cooling steam is supplied in the closed-circuit steam-cooling passages and compressor discharge air is supplied through the bore and into the wheel space cavities to cool the rotor.

Advanced Steels for Accident Tolerant Fuel Cladding in Current Light Water Reactors

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

After the March 2011 Fukushima events, the U.S. Congress directed the Department of Energy (DOE) to focus efforts on the development of fuel cladding materials with enhanced accident tolerance. In comparison with the stand-ard UO2-Zirconium based system, the new fuels need to tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operation conditions. Advanced steels such as iron-chromium-aluminum (FeCrAl) alloys are being investigated for degradation behavior both under normal operation conditions in high temperature water (e.g. 288°C) and under accident conditions for reaction with steam up to 1400°C. Commercial and experimental alloys were tested for several periods of time in 100% superheated steam from 800°C to 1475°C. Results show that FeCrAl alloys significantly outperform the resistance in steam of the current zirconium alloys.

Integrated hydrocarbon reforming system and controls

DOEpatents

Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Thijssen, Johannes; Davis, Robert; Papile, Christopher; Rumsey, Jennifer W.; Longo, Nathan; Cross, III, James C.; Rizzo, Vincent; Kleeburg, Gunther; Rindone, Michael; Block, Stephen G.; Sun, Maria; Morriseau, Brian D.; Hagan, Mark R.; Bowers, Brian

2003-11-04

A hydrocarbon reformer system including a first reactor configured to generate hydrogen-rich reformate by carrying out at least one of a non-catalytic thermal partial oxidation, a catalytic partial oxidation, a steam reforming, and any combinations thereof, a second reactor in fluid communication with the first reactor to receive the hydrogen-rich reformate, and having a catalyst for promoting a water gas shift reaction in the hydrogen-rich reformate, and a heat exchanger having a first mass of two-phase water therein and configured to exchange heat between the two-phase water and the hydrogen-rich reformate in the second reactor, the heat exchanger being in fluid communication with the first reactor so as to supply steam to the first reactor as a reactant is disclosed. The disclosed reformer includes an auxiliary reactor configured to generate heated water/steam and being in fluid communication with the heat exchanger of the second reactor to supply the heated water/steam to the heat exchanger.

Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2014-10-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC)more » system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.« less

REFLECTOR CONTROL OF A BOILING-WATER REACTOR

DOEpatents

Treshow, M.

1962-05-22

A line connecting the reactor with a spent steam condenser contains a valve set to open when the pressure in the reactor exceeds a predetermined value and an orifice on the upstream side of the valve. Another line connects the reflector with this line between the orifice and the valve. An excess steam pressure causes the valve to open and the flow of steam through the line draws water out of the reflector. Provision is also made for adding water to the reflector when the steam pressure drops. (AEC)

Dismantlement of the TSF-SNAP Reactor Assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peretz, Fred J

2009-01-01

This paper describes the dismantlement of the Tower Shielding Facility (TSF)?Systems for Nuclear Auxiliary Power (SNAP) reactor, a SNAP-10A reactor used to validate radiation source terms and shield performance models at Oak Ridge National Laboratory (ORNL) from 1967 through 1973. After shutdown, it was placed in storage at the Y-12 National Security Complex (Y-12), eventually falling under the auspices of the Highly Enriched Uranium (HEU) Disposition Program. To facilitate downblending of the HEU present in the fuel elements, the TSF-SNAP was moved to ORNL on June 24, 2006. The reactor assembly was removed from its packaging, inspected, and the sodium-potassiummore » (NaK) coolant was drained. A superheated steam process was used to chemically react the residual NaK inside the reactor assembly. The heat exchanger assembly was removed from the top of the reactor vessel, and the criticality safety sleeve was exchanged for a new safety sleeve that allowed for the removal of the vessel lid. A chain-mounted tubing cutter was used to separate the lid from the vessel, and the 36 fuel elements were removed and packaged in four U.S. Department of Transportation 2R/6M containers. The fuel elements were returned to Y-12 on July 13, 2006. The return of the fuel elements and disposal of all other reactor materials accomplished the formal objectives of the dismantlement project. In addition, a project model was established for the handling of a fully fueled liquid-metal?cooled reactor assembly. Current criticality safety codes have been benchmarked against experiments performed by Atomics International in the 1950s and 1960s. Execution of this project provides valuable experience applicable to future projects addressing space and liquid-metal-cooled reactors.« less

Heat-transfer analysis of double-pipe heat exchangers for indirect-cycle SCW NPP

NASA Astrophysics Data System (ADS)

Thind, Harwinder

SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. SuperCritical Water (SCW) Nuclear Power Plants (NPPs) are expected to have much higher operating parameters compared to current NPPs, i.e., pressure of about 25 MPa and outlet temperature up to 625 °C. This study presents the heat transfer analysis of an intermediate Heat exchanger (HX) design for indirect-cycle concepts of Pressure-Tube (PT) and Pressure-Vessel (PV) SCWRs. Thermodynamic configurations with an intermediate HX gives a possibility to have a single-reheat option for PT and PV SCWRs without introducing steam-reheat channels into a reactor. Similar to the current CANDU and Pressurized Water Reactor (PWR) NPPs, steam generators separate the primary loop from the secondary loop. In this way, the primary loop can be completely enclosed in a reactor containment building. This study analyzes the heat transfer from a SCW primary (reactor) loop to a SCW and Super-Heated Steam (SHS) secondary (turbine) loop using a double-pipe intermediate HX. The numerical model is developed with MATLAB and NIST REFPROP software. Water from the primary loop flows through the inner pipe, and water from the secondary loop flows through the annulus in the counter direction of the double-pipe HX. The analysis on the double-pipe HX shows temperature and profiles of thermophysical properties along the heated length of the HX. It was found that the pseudocritical region has a significant effect on the temperature profiles and heat-transfer area of the HX. An analysis shows the effect of variation in pressure, temperature, mass flow rate, and pipe size on the pseudocritical region and the heat-transfer area of the HX. The results from the numerical model can be used to optimize the heat-transfer area of the HX. The higher pressure difference on the hot side and higher temperature difference between the hot and cold sides reduces the pseudocritical-region length, thus decreases the heat-transfer surface area of the HX.

Experimental study of condensate subcooling with the use of a model of an air-cooled condenser

NASA Astrophysics Data System (ADS)

Sukhanov, V. A.; Bezukhov, A. P.; Bogov, I. A.; Dontsov, N. Y.; Volkovitsky, I. D.; Tolmachev, V. V.

2016-01-01

Water-supply deficit is now felt in many regions of the world. This hampers the construction of new steam-turbine and combined steam-and-gas thermal power plants. The use of dry cooling systems and, specifically, steam-turbine air-cooled condensers (ACCs) expands the choice of sites for the construction of such power plants. The significance of condensate subcooling Δ t as a parameter that negatively affects the engineering and economic performance of steam-turbine plants is thereby increased. The operation and design factors that influence the condensate subcooling in ACCs are revealed, and the research objective is, thus, formulated properly. The indicated research was conducted through physical modeling with the use of the Steam-Turbine Air-Cooled Condenser Unit specialized, multipurpose, laboratory bench. The design and the combined schematic and measurement diagram of this test bench are discussed. The experimental results are presented in the form of graphic dependences of the condensate subcooling value on cooling ratio m and relative weight content ɛ' of air in steam at the ACC inlet at different temperatures of cooling air t ca ' . The typical ranges of condensate subcooling variation (4 ≤ Δ t ≤ 6°C, 2 ≤ Δ t ≤ 4°C, and 0 ≤ Δ t ≤ 2°C) are identified based on the results of analysis of the attained Δ t levels in the ACC and numerous Δ t reduction estimates. The corresponding ranges of cooling ratio variation at different temperatures of cooling air at the ACC inlet are specified. The guidelines for choosing the adjusted ranges of cooling ratio variation with account of the results of experimental studies of the dependences of the absolute pressure of the steam-air mixture in the top header of the ACC and the heat flux density on the cooling ratio at different temperatures of cooling air at the ACC inlet are given.

Water cooled steam jet

DOEpatents

Wagner, Jr., Edward P.

1999-01-01

A water cooled steam jet for transferring fluid and preventing vapor lock, or vaporization of the fluid being transferred, has a venturi nozzle and a cooling jacket. The venturi nozzle produces a high velocity flow which creates a vacuum to draw fluid from a source of fluid. The venturi nozzle has a converging section connected to a source of steam, a diffuser section attached to an outlet and a throat portion disposed therebetween. The cooling jacket surrounds the venturi nozzle and a suction tube through which the fluid is being drawn into the venturi nozzle. Coolant flows through the cooling jacket. The cooling jacket dissipates heat generated by the venturi nozzle to prevent vapor lock.

Proposed Performance Evaluation Acceptance Test for Heat Recovery Incinerators

DTIC Science & Technology

1988-08-01

steam and the cooling water (if used). = Qye + Qwe = Mass flow of steam or water x enthalpy change. Qye = Wye x (hout - hin) Qwe = Wwe x (hout - hin...cooling water (if used). = Qye + Qwe = . Mass flow of steam or water x enthalpy change. Qye = Wye x (hout - hin) Qwe = Wwe x (hout - hin) = Wwe x (tout...transferred to recovery liquid (e.g., steam) Btu/hr 0.293 W Qwe Heat in water (cooling or Btu/hr 0.293 W quench) r Waste - S Sulfur lb/lb - kg/kg t

HORIZONTAL BOILING REACTOR SYSTEM

DOEpatents

Treshow, M.

1958-11-18

Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

Gas turbine combustor transition

DOEpatents

Coslow, Billy Joe; Whidden, Graydon Lane

1999-01-01

A method of converting a steam cooled transition to an air cooled transition in a gas turbine having a compressor in fluid communication with a combustor, a turbine section in fluid communication with the combustor, the transition disposed in a combustor shell and having a cooling circuit connecting a steam outlet and a steam inlet and wherein hot gas flows from the combustor through the transition and to the turbine section, includes forming an air outlet in the transition in fluid communication with the cooling circuit and providing for an air inlet in the transition in fluid communication with the cooling circuit.

Gas turbine combustor transition

DOEpatents

Coslow, B.J.; Whidden, G.L.

1999-05-25

A method is described for converting a steam cooled transition to an air cooled transition in a gas turbine having a compressor in fluid communication with a combustor, a turbine section in fluid communication with the combustor, the transition disposed in a combustor shell and having a cooling circuit connecting a steam outlet and a steam inlet and wherein hot gas flows from the combustor through the transition and to the turbine section, includes forming an air outlet in the transition in fluid communication with the cooling circuit and providing for an air inlet in the transition in fluid communication with the cooling circuit. 7 figs.

Methods for disassembling, replacing and assembling parts of a steam cooling system for a gas turbine

DOEpatents

Wilson, Ian D.; Wesorick, Ronald R.

2002-01-01

The steam cooling circuit for a gas turbine includes a bore tube assembly supplying steam to circumferentially spaced radial tubes coupled to supply elbows for transitioning the radial steam flow in an axial direction along steam supply tubes adjacent the rim of the rotor. The supply tubes supply steam to circumferentially spaced manifold segments located on the aft side of the 1-2 spacer for supplying steam to the buckets of the first and second stages. Spent return steam from these buckets flows to a plurality of circumferentially spaced return manifold segments disposed on the forward face of the 1-2 spacer. Crossover tubes couple the steam supply from the steam supply manifold segments through the 1-2 spacer to the buckets of the first stage. Crossover tubes through the 1-2 spacer also return steam from the buckets of the second stage to the return manifold segments. Axially extending return tubes convey spent cooling steam from the return manifold segments to radial tubes via return elbows. The bore tube assembly, radial tubes, elbows, manifold segments and crossover tubes are removable from the turbine rotor and replaceable.

Cooling of core debris and the impact on containment pressure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, J.W.

1981-07-01

An evaluation of the core debris/water interactions associated with a postulated meltdown of a PWR and its impact on the containment pressure is presented. In the event of a complete core meltdown in a PWR, the interaction of molten debris with water in the bottom head of the reactor vessel could result in complete evaporation of water and breach of the vessel wall. In the reactor cavity, the debris-water interaction may lead to a rapid generation of steam, which could lead to pressures beyond the containment building limit. Previous analysis of the debris-water interactions with the MARCH code was basedmore » on the single-sphere model, in which the internal and surface heat transfer are the controlling mechanisms. In this study, the potential in-vessel and ex-vessel debris-water interactions are analyzed in terms of porous debris bed models. The debris cooling and steam generation are controlled by the hydrodynamics of the two-phase flow. The porous models developed by Dhir-Catton and by Lipinski were examined and used to test their impact on containment dynamics. The tests include several particle sizes from 1 mm to 50 mm. Detailed transient data on the pressure, temperature, and mass of steam in the containment building was obtained for all cases. Bands of pressure variation which represents the possible pressure rise under accident conditions were obtained for the Dhir-Catton model and for the Lipinski model. The results show that, for the case of a wet cavity, the magnitude of the predicted pressure rises is not strongly affected by the different models. The occurrence of the peak pressure, however, is considerably delayed by using the debris bed model. For the case of a dry cavity, a large reduction of the peak pressure is obtained by using the debris bed model.« less

Water cooled steam jet

DOEpatents

Wagner, E.P. Jr.

1999-01-12

A water cooled steam jet for transferring fluid and preventing vapor lock, or vaporization of the fluid being transferred, has a venturi nozzle and a cooling jacket. The venturi nozzle produces a high velocity flow which creates a vacuum to draw fluid from a source of fluid. The venturi nozzle has a converging section connected to a source of steam, a diffuser section attached to an outlet and a throat portion disposed there between. The cooling jacket surrounds the venturi nozzle and a suction tube through which the fluid is being drawn into the venturi nozzle. Coolant flows through the cooling jacket. The cooling jacket dissipates heat generated by the venturi nozzle to prevent vapor lock. 2 figs.

Advances of zeolite based membrane for hydrogen production via water gas shift reaction

NASA Astrophysics Data System (ADS)

Makertihartha, I. G. B. N.; Zunita, M.; Rizki, Z.; Dharmawijaya, P. T.

2017-07-01

Hydrogen is considered as a promising energy vector which can be obtained from various renewable sources. However, an efficient hydrogen production technology is still challenging. One technology to produce hydrogen with very high capacity with low cost is through water gas shift (WGS) reaction. Water gas shift reaction is an equilibrium reaction that produces hydrogen from syngas mixture by the introduction of steam. Conventional WGS reaction employs two or more reactors in series with inter-cooling to maximize conversion for a given volume of catalyst. Membrane reactor as new technology can cope several drawbacks of conventional reactor by removing reaction product and the reaction will favour towards product formation. Zeolite has properties namely high temperature, chemical resistant, and low price makes it suitable for membrane reactor applications. Moreover, it has been employed for years as hydrogen selective layer. This review paper is focusing on the development of membrane reactor for efficient water gas shift reaction to produce high purity hydrogen and carbon dioxide. Development of membrane reactor is discussed further related to its modification towards efficient reaction and separation from WGS reaction mixture. Moreover, zeolite framework suitable for WGS membrane reactor will be discussed more deeply.

Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Philip E. MacDonald

2005-01-01

The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current Light Water Reactors [LWRs]) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus, the need for a pressurizer, steam generators, steam separators, and dryers is eliminated. The main mission ofmore » the SCWR is generation of low-cost electricity. It is built upon two proven technologies: LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which are also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa, with core inlet and outlet temperatures of 280 and 500 C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel, to then flow down through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was organized into three tasks: Task 1. Fuel-cycle Neutronic Analysis and Reactor Core Design Task 2. Fuel Cladding and Structural Material Corrosion and Stress Corrosion Cracking Task 3. Plant Engineering and Reactor Safety Analysis. moderator rods. materials.« less

Steam generator for liquid metal fast breeder reactor

DOEpatents

Gillett, James E.; Garner, Daniel C.; Wineman, Arthur L.; Robey, Robert M.

1985-01-01

Improvements in the design of internal components of J-shaped steam generators for liquid metal fast breeder reactors. Complex design improvements have been made to the internals of J-shaped steam generators which improvements are intended to reduce tube vibration, tube jamming, flow problems in the upper portion of the steam generator, manufacturing complexities in tube spacer attachments, thermal stripping potentials and difficulties in the weld fabrication of certain components.

Decommissioning of German Nuclear Research Facilities under the Governance of the Federal Ministry of Education and Research

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weigl, M.

2008-07-01

Since the announcement of the first nuclear program in 1956, nuclear R and D in Germany has been supported by the Federal Government under four nuclear programs and later on under more general energy R and D programs. The original goal was to help German industry to achieve safe, low-cost generation of energy and self-sufficiency in the various branches of nuclear technology, including the fast breeder reactor and the fuel cycle. Several national research centers were established to host or operate experimental and demonstration plants. These are mainly located at the sites of the national research centers at Juelich andmore » Karlsruhe. In the meantime, all these facilities were shut down and most of them are now in a state of decommissioning and dismantling (D and D). Meanwhile, Germany is one of the leading countries in the world in the field of D and D. Two big demonstration plants, the Niederaichbach Nuclear Power Plant (KKN) a heavy-water cooled pressure tube reactor with carbon-dioxide cooling and the Karlstein Superheated Steam Reactor (HDR) a boiling light water reactor with a thermal power of 100 MW, are totally dismantled and 'green field' is reached. For two other projects the return to 'green field' sites will be reached by the end of this decade. These are the dismantling of the Multi-Purpose Research Reactor (MZFR) and the Compact Sodium Cooled Reactor (KNK) both located at the Forschungszentrum Karlsruhe. Within these projects a lot of new solutions und innovative techniques were tested, which were developed at German universities and in small and medium sized companies mostly funded by the Federal Ministry of Education and Research (BMBF). For example, high performance underwater cutting technologies like plasma arc cutting and contact arc metal cutting. (authors)« less

Core cooling under accident conditions at the high flux beam reactor (HFBR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tichler, P.; Cheng, L.; Fauske, H.

In certain accident scenarios, e.g. loss of coolant accidents (LOCA) all forced flow cooling is lost. Decay heating causes a temperature increase in the core coolant and the resulting thermal buoyancy causes a reversal of the flow direction to a natural circulation mode. Although there was experimental evidence during the reactor design period (1958--1963) that the heat removal capacity in the fully developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. In a LOCA scenario where even limited fuelmore » damage occurs and natural circulation is established, fission product gases could be carried from the damaged fuel by steam into areas where operator access is required to maintain the core in a coolable configuration. This would force evacuation of the building and lead to extensive core damage. As a result the HFBR was shut down by the Department of Energy (DOE) and an extensive review of the HFBR was initiated. In an effort to address this issue BNL developed a model designed to predict the heat removal limit during flow reversal that was found to be in good agreement with the test results. Currently a thermal-hydraulic test program is being developed to provide a more realistic and defensible estimate of the flow reversal heat removal limit so that the reactor power level can be increased.« less

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.

Closed loop steam cooled airfoil

DOEpatents

Widrig, Scott M.; Rudolph, Ronald J.; Wagner, Gregg P.

2006-04-18

An airfoil, a method of manufacturing an airfoil, and a system for cooling an airfoil is provided. The cooling system can be used with an airfoil located in the first stages of a combustion turbine within a combined cycle power generation plant and involves flowing closed loop steam through a pin array set within an airfoil. The airfoil can comprise a cavity having a cooling chamber bounded by an interior wall and an exterior wall so that steam can enter the cavity, pass through the pin array, and then return to the cavity to thereby cool the airfoil. The method of manufacturing an airfoil can include a type of lost wax investment casting process in which a pin array is cast into an airfoil to form a cooling chamber.

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using themore » WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)« less

Core thermal response and hydrogen generation of the N Reactor hydrogen mitigation design basis accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, M.D.; Lombardo, N.J.; Heard, F.J.

1988-04-01

Calculations were performed to determine core heatup, core damage, and subsequent hydrogen production of a hypothetical loss-of-cooling accident at the Department of Energy's N Reactor. The thermal transient response of the reactor core was solved using the TRUMP-BD computer program. Estimates of whole-core thermal damage and hydrogen production were made by weighting the results of multiple half-length pressure tube simulations at various power levels. The Baker-Just and Wilson parabolic rate equations for the metal-water chemical reactions modeled the key phenomena of chemical energy and hydrogen evolution. Unlimited steam was assumed available for continuous oxidation of exposed Zircaloy-2 surfaces and formore » uranium metal with fuel cladding beyond the failure temperature (1038 C). Intact fuel geometry was modeled. Maximum fuel temperatures (1181 C) in the cooled central regions of the core were predicted to occur one-half hour into the accident scenario. Maximum fuel temperatures of 1447 C occurred in the core GSCS-regions at the end of the 10-h transient. After 10-h 26% of the fuel inventory was predicted to have failed. Peak hydrogen evolution equaled 42 g/s, while 10-h integrated hydrogen evolution equaled 167 kg. 12 refs., 12 figs., 2 tabs.« less

Heat transfer performance comparison of steam and air in gas turbine cooling channels with different rib angles

NASA Astrophysics Data System (ADS)

Shi, Xiaojun; Gao, Jianmin; Xu, Liang; Li, Fajin

2013-11-01

Using steam as working fluid to replace compressed air is a promising cooling technology for internal cooling passages of blades and vanes. The local heat transfer characteristics and the thermal performance of steam flow in wide aspect ratio channels ( W/ H = 2) with different angled ribs on two opposite walls have been experimentally investigated in this paper. The averaged Nusselt number ratios and the friction factor ratios of steam and air in four ribbed channels were also measured under the same test conditions for comparison. The Reynolds number range is 6,000-70,000. The rib angles are 90°, 60°, 45°, and 30°, respectively. The rib height to hydraulic diameter ratio is 0.047. The pitch-to-rib height ratio is 10. The results show that the Nusselt number ratios of steam are 1.19-1.32 times greater than those of air over the range of Reynolds numbers studied. For wide aspect ratio channels using steam as the coolant, the 60° angled ribs has the best heat transfer performance and is recommended for cooling design.

Safety and environmental aspects of organic coolants for fusion facilities

NASA Astrophysics Data System (ADS)

Natalizio, A.; Hollies, R. E.; Gierszewski, P.

1993-06-01

Organic coolants, such as OS-84, offer unique advantages for fusion reactor applications. These advantages are with respect to both reactor operation and safety. The key operational advantage is a coolant that can provide high temperature (350-400°C) at modest pressure (2-4 MPa). These temperatures are needed for conditioning the plasma-facing components and, in reactors, for achieving high thermodynamic conversion efficiencies (>40%). The key safety advantage of organic coolants is the low vapor pressure, which significantly reduces the containment pressurization transient (relative to water) following a loss of coolant event. Also, from an occupational dose viewpoint, organic coolants significantly reduce corrosion and erosion inside the cooling system and consequently reduce the quantity of activation products deposited in cooling system equipment. On the negative side, organic coolants undergo both pyrolytic and radiolytic decomposition, and are flammable. While the decomposition rate can be minimized by coolant system design (by reducing coolant inventories exposed to neutron flux and to high temperatures), decomposition products are formed and these degrade the coolant properties. Both heavy compounds and light gases are produced from the decomposition process, and both must be removed to maintain adequate coolant properties. As these hydrocarbons may become tritiated by permeation, or activated through impurities, their disposal could create an environmental concern. Because of this potential waste disposal problem, consideration has been given to the recycling of both the light and heavy products, thereby reducing the quantity of waste to be disposed. Preliminary assessments made for various fusion reactor designs, including ITER, suggest that it is feasible to use organic coolants for several applications. These applications range from first wall and blanket coolant (the most demanding with respect to decomposition), to shield and vacuum vessel cooling, to an intermediate cooling loop removing heat from a liquid metal loop and transferring it to a steam generator or heat exchanger.

Thermochemically recuperated and steam cooled gas turbine system

DOEpatents

Viscovich, Paul W.; Bannister, Ronald L.

1995-01-01

A gas turbine system in which the expanded gas from the turbine section is used to generate the steam in a heat recovery steam generator and to heat a mixture of gaseous hydrocarbon fuel and the steam in a reformer. The reformer converts the hydrocarbon gas to hydrogen and carbon monoxide for combustion in a combustor. A portion of the steam from the heat recovery steam generator is used to cool components, such as the stationary vanes, in the turbine section, thereby superheating the steam. The superheated steam is mixed into the hydrocarbon gas upstream of the reformer, thereby eliminating the need to raise the temperature of the expanded gas discharged from the turbine section in order to achieve effective conversion of the hydrocarbon gas.

Active cooling-based surface confinement system for thermal soil treatment

DOEpatents

Aines, R.D.; Newmark, R.L.

1997-10-28

A thermal barrier is disclosed for surface confinement with active cooling to control subsurface pressures during thermal remediation of shallow (5-20 feet) underground contaminants. If steam injection is used for underground heating, the actively cooled thermal barrier allows the steam to be injected into soil at pressures much higher (20-60 psi) than the confining strength of the soil, while preventing steam breakthrough. The rising steam is condensed to liquid water at the thermal barrier-ground surface interface. The rapid temperature drop forced by the thermal barrier drops the subsurface pressure to below atmospheric pressure. The steam and contaminant vapors are contained by the thermal blanket, which can be made of a variety of materials such as steel plates, concrete slabs, membranes, fabric bags, or rubber bladders. 1 fig.

Active cooling-based surface confinement system for thermal soil treatment

DOEpatents

Aines, Roger D.; Newmark, Robin L.

1997-01-01

A thermal barrier is disclosed for surface confinement with active cooling to control subsurface pressures during thermal remediation of shallow (5-20 feet) underground contaminants. If steam injection is used for underground heating, the actively cooled thermal barrier allows the steam to be injected into soil at pressures much higher (20-60 psi) than the confining strength of the soil, while preventing steam breakthrough. The rising steam is condensed to liquid water at the thermal barrier-ground surface interface. The rapid temperature drop forced by the thermal barrier drops the subsurface pressure to below atmospheric pressure. The steam and contaminant vapors are contained by the thermal blanket, which can be made of a variety of materials such as steel plates, concrete slabs, membranes, fabric bags, or rubber bladders.

System Evaluation and Economic Analysis of a HTGR Powered High-Temperature Electrolysis Hydrogen Production Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael G. McKellar; Edwin A. Harvego; Anastasia A. Gandrik

2010-10-01

A 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 power conversion unit will be a Rankine steam cycle with a power conversion efficiency of 40%. The reference hydrogen production plantmore » 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 40.4% at a hydrogen production rate of 1.75 kg/s and an oxygen production rate of 13.8 kg/s. An economic analysis of this plant was performed with 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 cost of $3.67/kg of hydrogen assuming an internal rate of return, IRR, of 12% and a debt to equity ratio of 80%/20%. A second analysis shows that if the power cycle efficiency increases to 44.4%, the hydrogen production efficiency increases to 42.8% and the hydrogen and oxygen production rates are 1.85 kg/s and 14.6 kg/s respectively. At the higher power cycle efficiency and an IRR of 12% the cost of hydrogen production is $3.50/kg.« less

Gas turbine row #1 steam cooled vane

DOEpatents

Cunha, Frank J.

2000-01-01

A design for a vane segment having a closed-loop steam cooling system is provided. The vane segment comprises an outer shroud, an inner shroud and an airfoil, each component having a target surface on the inside surface of its walls. A plurality of rectangular waffle structures are provided on the target surface to enhance heat transfer between each component and cooling steam. Channel systems are provided in the shrouds to improve the flow of steam through the shrouds. Insert legs located in cavities in the airfoil are also provided. Each insert leg comprises outer channels located on a perimeter of the leg, each outer channel having an outer wall and impingement holes on the outer wall for producing impingement jets of cooling steam to contact the airfoil's target surface. Each insert leg further comprises a plurality of substantially rectangular-shaped ribs located on the outer wall and a plurality of openings located between outer channels of the leg to minimize cross flow degradation.

Interfacial area transport of steam-water two-phase flow in a vertical annulus at elevated pressures

NASA Astrophysics Data System (ADS)

Ozar, Basar

Analysis of accident scenarios in nuclear reactors are done by using codes such as TRACE and RELAP5. Large oscillations in the core void fraction are observed in calculations of advanced passive light water reactors (ALWRs), especially during the low pressure long-term cooling phase. These oscillations are attributed to be numerical in nature and served to limit the accuracy as well as the credibility of the calculations. One of the root causes of these unphysical oscillations is determined to be flow regime transitions caused by the usage of static flow regime maps. The interfacial area transport equation was proposed earlier in order to address these issues. Previous research successfully developed the foundation of the interfacial area transport equation and the experimental techniques needed for the measurement of interfacial area, bubble diameters and velocities. In the past, an extensive database has been then generated for adiabatic air-water conditions in vertical upward and downward bubbly-churn turbulent flows in pipes. Using this database, mechanistic models for the creation (bubble breakup) and destruction (bubble coalescence) of interfacial area have been developed for the bubblyslug flow regime transition. However, none of these studies investigated the effect of phase change. To address this need, a heated annular test section was designed and constructed. The design relied on a three level scaling approach: geometric scaling; hydrodynamic scaling; thermal scaling. The test section consisted of a heated and unheated section in order to study the sub-cooled boiling and bulk condensation/flashing and evaporation phenomena, respectively. Steam-water two-phase flow tests were conducted under sub-cooled boiling conditions in the heated section and with sub-cooled/super-heated bulk liquid in the unheated section. The modeling of interfacial area transport equation with phase change effects was introduced and discussed. Constitutive relations, which took phase change effects into account, for interfacial area transport equation were proposed and implemented. Effects of these constitutive relations on the prediction capability of the transport equation were discussed.

Oxygen transport membrane reactor based method and system for generating electric power

DOEpatents

Kelly, Sean M.; Chakravarti, Shrikar; Li, Juan

2017-02-07

A carbon capture enabled system and method for generating electric power and/or fuel from methane containing sources using oxygen transport membranes by first converting the methane containing feed gas into a high pressure synthesis gas. Then, in one configuration the synthesis gas is combusted in oxy-combustion mode in oxygen transport membranes based boiler reactor operating at a pressure at least twice that of ambient pressure and the heat generated heats steam in thermally coupled steam generation tubes within the boiler reactor; the steam is expanded in steam turbine to generate power; and the carbon dioxide rich effluent leaving the boiler reactor is processed to isolate carbon. In another configuration the synthesis gas is further treated in a gas conditioning system configured for carbon capture in a pre-combustion mode using water gas shift reactors and acid gas removal units to produce hydrogen or hydrogen-rich fuel gas that fuels an integrated gas turbine and steam turbine system to generate power. The disclosed method and system can also be adapted to integrate with coal gasification systems to produce power from both coal and methane containing sources with greater than 90% carbon isolation.

Cooling system for a gas turbine

DOEpatents

Wilson, Ian David; Salamah, Samir Armando; Bylina, Noel Jacob

2003-01-01

A plurality of arcuate circumferentially spaced supply and return manifold segments are arranged on the rim of a rotor for respectively receiving and distributing cooling steam through exit ports for distribution to first and second-stage buckets and receiving spent cooling steam from the first and second-stage buckets through inlet ports for transmission to axially extending return passages. Each of the supply and return manifold segments has a retention system for precluding substantial axial, radial and circumferential displacement relative to the rotor. The segments also include guide vanes for minimizing pressure losses in the supply and return of the cooling steam. The segments lie substantially equal distances from the centerline of the rotor and crossover tubes extend through each of the segments for communicating steam between the axially adjacent buckets of the first and second stages, respectively.

Studies of the use of high-temperature nuclear heat from an HTGR for hydrogen production

NASA Technical Reports Server (NTRS)

Peterman, D. D.; Fontaine, R. W.; Quade, R. N.; Halvers, L. J.; Jahromi, A. M.

1975-01-01

The results of a study which surveyed various methods of hydrogen production using nuclear and fossil energy are presented. A description of these methods is provided, and efficiencies are calculated for each case. The process designs of systems that utilize the heat from a general atomic high temperature gas cooled reactor with a steam methane reformer and feed the reformer with substitute natural gas manufactured from coal, using reforming temperatures, are presented. The capital costs for these systems and the resultant hydrogen production price for these cases are discussed along with a research and development program.

Thermodynamic Analysis of the Use a Chemical Heat Pump to Link a Supercritical Water-Cooled Nuclear Reactor and a Thermochemical Water-Splitting Cycle for Hydrogen Production

NASA Astrophysics Data System (ADS)

Granovskii, Mikhail; Dincer, Ibrahim; Rosen, Marc A.; Pioro, Igor

Increases in the power generation efficiency of nuclear power plants (NPPs) are mainly limited by the permissible temperatures in nuclear reactors and the corresponding temperatures and pressures of the coolants in reactors. Coolant parameters are limited by the corrosion rates of materials and nuclear-reactor safety constraints. The advanced construction materials for the next generation of CANDU reactors, which employ supercritical water (SCW) as a coolant and heat carrier, permit improved “steam” parameters (outlet temperatures up to 625°C and pressures of about 25 MPa). An increase in the temperature of steam allows it to be utilized in thermochemical water splitting cycles to produce hydrogen. These methods are considered by many to be among the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require an intensive heat supply at temperatures higher than 550-600°C. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump, which increases the temperature of the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. Here, a high-temperature chemical heat pump, which employs the reversible catalytic methane conversion reaction, is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with the second steam cycle of a SCW nuclear power generation plant on one side and a thermochemical water splitting cycle on the other, increases the temperature of the “nuclear” heat and, consequently, the intensity of heat transfer into the water splitting cycle. A comparative preliminary thermodynamic analysis is conducted of the combined system comprising a SCW nuclear power generation plant and a chemical heat pump, which provides high-temperature heat to a thermochemical water splitting cycle for hydrogen production. It is concluded that the proposed chemical heat pump permits the utilization efficiency of nuclear energy to be improved by at least 2% without jeopardizing nuclear reactor safety. Based on this analysis, further research appears to be merited on the proposed advanced design of a nuclear power generation plant combined with a chemical heat pump, and implementation in appropriate applications seems worthwhile.

Retention sleeve for a thermal medium carrying tube in a gas turbine

DOEpatents

Lathrop, Norman Douglas; Czachor, Robert Paul

2003-01-01

Multiple tubes are connected to steam supply and spent cooling steam return manifolds for supplying cooling steam to buckets and returning spent cooling steam from the buckets to the manifolds, respectively. The tubes are prevented from axial movement in one direction by flanges engaging end faces of the spacer between the first and second-stage wheels. Retention sleeves are disposed about cantilevered ends of the tubes. One end of the retention sleeve engages an enlarged flange on the tube, while an opposite end is spaced axially from an end face of the adjoining wheel, forming a gap, enabling thermal expansion of the tubes and limiting axial displacement of the tube in the opposite direction.

Film cooling air pocket in a closed loop cooled airfoil

DOEpatents

Yu, Yufeng Phillip; Itzel, Gary Michael; Osgood, Sarah Jane; Bagepalli, Radhakrishna; Webbon, Waylon Willard; Burdgick, Steven Sebastian

2002-01-01

Turbine stator vane segments have radially inner and outer walls with vanes extending between them. The inner and outer walls are compartmentalized and have impingement plates. Steam flowing into the outer wall plenum passes through the impingement plate for impingement cooling of the outer wall upper surface. The spent impingement steam flows into cavities of the vane having inserts for impingement cooling the walls of the vane. The steam passes into the inner wall and through the impingement plate for impingement cooling of the inner wall surface and for return through return cavities having inserts for impingement cooling of the vane surfaces. To provide for air film cooing of select portions of the airfoil outer surface, at least one air pocket is defined on a wall of at least one of the cavities. Each air pocket is substantially closed with respect to the cooling medium in the cavity and cooling air pumped to the air pocket flows through outlet apertures in the wall of the airfoil to cool the same.

Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.

1984-06-01

ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Componentmore » Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.« less

Thermochemically recuperated and steam cooled gas turbine system

DOEpatents

Viscovich, P.W.; Bannister, R.L.

1995-07-11

A gas turbine system is described in which the expanded gas from the turbine section is used to generate the steam in a heat recovery steam generator and to heat a mixture of gaseous hydrocarbon fuel and the steam in a reformer. The reformer converts the hydrocarbon gas to hydrogen and carbon monoxide for combustion in a combustor. A portion of the steam from the heat recovery steam generator is used to cool components, such as the stationary vanes, in the turbine section, thereby superheating the steam. The superheated steam is mixed into the hydrocarbon gas upstream of the reformer, thereby eliminating the need to raise the temperature of the expanded gas discharged from the turbine section in order to achieve effective conversion of the hydrocarbon gas. 4 figs.

Method for generating hydrogen for fuel cells

DOEpatents

Ahmed, Shabbir; Lee, Sheldon H. D.; Carter, John David; Krumpelt, Michael

2004-03-30

A method of producing a H.sub.2 rich gas stream includes supplying an O.sub.2 rich gas, steam, and fuel to an inner reforming zone of a fuel processor that includes a partial oxidation catalyst and a steam reforming catalyst or a combined partial oxidation and stream reforming catalyst. The method also includes contacting the O.sub.2 rich gas, steam, and fuel with the partial oxidation catalyst and the steam reforming catalyst or the combined partial oxidation and stream reforming catalyst in the inner reforming zone to generate a hot reformate stream. The method still further includes cooling the hot reformate stream in a cooling zone to produce a cooled reformate stream. Additionally, the method includes removing sulfur-containing compounds from the cooled reformate stream by contacting the cooled reformate stream with a sulfur removal agent. The method still further includes contacting the cooled reformate stream with a catalyst that converts water and carbon monoxide to carbon dioxide and H.sub.2 in a water-gas-shift zone to produce a final reformate stream in the fuel processor.

Fuel processor and method for generating hydrogen for fuel cells

DOEpatents

Ahmed, Shabbir [Naperville, IL; Lee, Sheldon H. D. [Willowbrook, IL; Carter, John David [Bolingbrook, IL; Krumpelt, Michael [Naperville, IL; Myers, Deborah J [Lisle, IL

2009-07-21

A method of producing a H.sub.2 rich gas stream includes supplying an O.sub.2 rich gas, steam, and fuel to an inner reforming zone of a fuel processor that includes a partial oxidation catalyst and a steam reforming catalyst or a combined partial oxidation and stream reforming catalyst. The method also includes contacting the O.sub.2 rich gas, steam, and fuel with the partial oxidation catalyst and the steam reforming catalyst or the combined partial oxidation and stream reforming catalyst in the inner reforming zone to generate a hot reformate stream. The method still further includes cooling the hot reformate stream in a cooling zone to produce a cooled reformate stream. Additionally, the method includes removing sulfur-containing compounds from the cooled reformate stream by contacting the cooled reformate stream with a sulfur removal agent. The method still further includes contacting the cooled reformate stream with a catalyst that converts water and carbon monoxide to carbon dioxide and H.sub.2 in a water-gas-shift zone to produce a final reformate stream in the fuel processor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leskovar, Matjaz; Koncar, Bostjan

An ex-vessel steam explosion may occur when during a severe reactor accident the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles at later times, during the expansion of the highly pressurized water vapor, that may endanger surrounding structures. In contrast to specialized steammore » explosion CFD codes, where the steam explosion is modeled on micro-scale using fundamental averaged multiphase flow conservation equations, in the presented approach the steam explosion is modeled in a simplified manner as an expanding high-pressure pre-mixture of dispersed molten fuel, liquid water and vapor. Applying the developed steam explosion model, a comprehensive analysis of the ex-vessel steam explosion in a typical PWR reactor cavity was done using the CFD code CFX-10. At four selected locations, which are of importance for the assessment of the vulnerability of cavity structures, the pressure histories were recorded and the corresponding pressure impulses calculated. The pressure impulses determine the destructive potential of the steam explosion and represent the input for the structural mechanical analysis of the cavity structures. The simulation results show that the pressure impulses depend mainly on the steam explosion energy conversion ratio, whereas the influence of the pre-mixture vapor volume fraction, which is a parameter in our model and determines the maximum steam explosion pressure, is not significant. (authors)« less

The nuclear battery

NASA Astrophysics Data System (ADS)

Kozier, K. S.; Rosinger, H. E.

The evolution and present status of an Atomic Energy of Canada Limited program to develop a small, solid-state, passively cooled reactor power supply known as the Nuclear Battery is reviewed. Key technical features of the Nuclear Battery reactor core include a heat-pipe primary heat transport system, graphite neutron moderator, low-enriched uranium TRISO coated-particle fuel and the use of burnable poisons for long-term reactivity control. An external secondary heat transport system extracts useful heat energy, which may be converted into electricity in an organic Rankine cycle engine or used to produce high-pressure steam. The present reference design is capable of producing about 2400 kW(t) (about 600 kW(e) net) for 15 full-power years. Technical and safety features are described along with recent progress in component hardware development programs and market assessment work.

Thermal and hydraulic analysis of a cylindrical blanket module design for a tokamak reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, A.Y.

1978-10-01

Various existing blanket design concepts for a tokamak fusion reactor were evaluated and assessed. These included the demonstration power reactors of ORNL, GA and others. As a result of this study, a cylindrical, modularized blanket design concept was developed. The module is a double-walled, stainless steel 316 cylinder containing liquid lithium for tritium breeding and is cooled by pressurized helium. Steady state and transient thermal conditions under normal and some off-design conditions were analyzed and presented. At the steady state reference operating point the maximum structure temperature is 452/sup 0/C at the maximum stressed location and is 495/sup 0/C atmore » the less stressed location. The coolant inlet pressure is 54.4 atm, the inlet temperature is 200/sup 0/C and the exit temperature is 435/sup 0/C. The coolant could be utilized with a helium/steam turbine power conversion system with a cycle thermal efficiency of 30.8%.« less

Westinghouse Small Modular Reactor nuclear steam supply system design

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 andmore » 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 generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)« less

Experimental Investigations in a Reactor Cavity Cooling System with Advanced Instrumentation for the Study of Instabilities, Oscillations, and Transients

NASA Astrophysics Data System (ADS)

Tompkins, Casey A.

A research team at University of Wisconsin - Madison designed and constructed a 1/4 height scaled experimental facility to study two-phase natural circulation cooling in a water-based reactor cavity cooling system (WRCCS) for decay heat removal in an advanced high temperature reactor. The facility is capable of natural circulation operation scaled for simulated decay heat removal (up to 28.5 kW m-2 (45 kW) input power, which is equivalent to 14.25 kW m-2 (6.8 MW) at full scale) and pressurized up to 2 bar. The UW-WRCCS facility has been used to study instabilities and oscillations observed during natural circulation flow due to evaporation of the water inventory. During two-phase operation, the system exhibits flow oscillations and excursions, which cause thermal oscillations in the structure. This can cause degradation in the mechanical structure at welds and limit heat transfer to the coolant. The facility is equipped with wire mesh sensors (WMS) that enable high-resolution measurements of the void fraction and steam velocities in order to study the instability's and oscillation's growth and decay during transient operation. Multiple perturbations to the system's operating point in pressure and inlet throttling have shown that the oscillatory behavior present under normal two-phase operating conditions can be damped and removed. Furthermore, with steady-state modeling it was discovered that a flow regime transition instability is the primary cause of oscillations in the UW-WRCCS facility under unperturbed conditions and that proper orifice selection can move the system into a stable operating regime.

Film cooling for a closed loop cooled airfoil

DOEpatents

Burdgick, Steven Sebastian; Yu, Yufeng Phillip; Itzel, Gary Michael

2003-01-01

Turbine stator vane segments have radially inner and outer walls with vanes extending therebetween. The inner and outer walls are compartmentalized and have impingement plates. Steam flowing into the outer wall plenum passes through the impingement plate for impingement cooling of the outer wall upper surface. The spent impingement steam flows into cavities of the vane having inserts for impingement cooling the walls of the vane. The steam passes into the inner wall and through the impingement plate for impingement cooling of the inner wall surface and for return through return cavities having inserts for impingement cooling of the vane surfaces. At least one film cooling hole is defined through a wall of at least one of the cavities for flow communication between an interior of the cavity and an exterior of the vane. The film cooling hole(s) are defined adjacent a potential low LCF life region, so that cooling medium that bleeds out through the film cooling hole(s) reduces a thermal gradient in a vicinity thereof, thereby the increase the LCF life of that region.

Development of a test device to characterize thermal protective performance of fabrics against hot steam and thermal radiation

NASA Astrophysics Data System (ADS)

Su, Yun; Li, Jun

2016-12-01

Steam burns severely threaten the life of firefighters in the course of their fire-ground activities. The aim of this paper was to characterize thermal protective performance of flame-retardant fabrics exposed to hot steam and low-level thermal radiation. An improved testing apparatus based on ASTM F2731-11 was developed in order to simulate the routine fire-ground conditions by controlling steam pressure, flow rate and temperature of steam box. The thermal protective performance of single-layer and multi-layer fabric system with/without an air gap was studied based on the calibrated tester. It was indicated that the new testing apparatus effectively evaluated thermal properties of fabric in hot steam and thermal radiation. Hot steam significantly exacerbated the skin burn injuries while the condensed water on the skin’s surface contributed to cool down the skin tissues during the cooling. Also, the absorbed thermal energy during the exposure and the cooling was mainly determined by the fabric’s configuration, the air gap size, the exposure time and the existence of hot steam. The research provides a effective method to characterize the thermal protection of fabric in complex conditions, which will help in optimization of thermal protection performance of clothing and reduction of steam burn.

Thermal analysis of cylindrical natural-gas steam reformer for 5 kW PEMFC

NASA Astrophysics Data System (ADS)

Jo, Taehyun; Han, Junhee; Koo, Bonchan; Lee, Dohyung

2016-11-01

The thermal characteristics of a natural-gas based cylindrical steam reformer coupled with a combustor are investigated for the use with a 5 kW polymer electrolyte membrane fuel cell. A reactor unit equipped with nickel-based catalysts was designed to activate the steam reforming reaction without the inclusion of high-temperature shift and low-temperature shift processes. Reactor temperature distribution and its overall thermal efficiency depend on various inlet conditions such as the equivalence ratio, the steam to carbon ratio (SCR), and the fuel distribution ratio (FDR) into the reactor and the combustor components. These experiments attempted to analyze the reformer's thermal and chemical properties through quantitative evaluation of product composition and heat exchange between the combustor and the reactor. FDR is critical factor in determining the overall performance as unbalanced fuel injection into the reactor and the combustor deteriorates overall thermal efficiency. Local temperature distribution also influences greatly on the fuel conversion rate and thermal efficiency. For the experiments, the operation conditions were set as SCR was in range of 2.5-4.0 and FDR was in 0.4-0.7 along with equivalence ratio of 0.9-1.1; optimum results were observed for FDR of 0.63 and SCR of 3.0 in the cylindrical steam reformer.

Co-Production of Electricity and Hydrogen Using a Novel Iron-based Catalyst

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hilaly, Ahmad; Georgas, Adam; Leboreiro, Jose

2011-09-30

The primary objective of this project was to develop a hydrogen production technology for gasification applications based on a circulating fluid-bed reactor and an attrition resistant iron catalyst. The work towards achieving this objective consisted of three key activities: Development of an iron-based catalyst suitable for a circulating fluid-bed reactor; Design, construction, and operation of a bench-scale circulating fluid-bed reactor system for hydrogen production; Techno-economic analysis of the steam-iron and the pressure swing adsorption hydrogen production processes. This report describes the work completed in each of these activities during this project. The catalyst development and testing program prepared and iron-basedmore » catalysts using different support and promoters to identify catalysts that had sufficient activity for cyclic reduction with syngas and steam oxidation and attrition resistance to enable use in a circulating fluid-bed reactor system. The best performing catalyst from this catalyst development program was produced by a commercial catalyst toll manufacturer to support the bench-scale testing activities. The reactor testing systems used during material development evaluated catalysts in a single fluid-bed reactor by cycling between reduction with syngas and oxidation with steam. The prototype SIP reactor system (PSRS) consisted of two circulating fluid-bed reactors with the iron catalyst being transferred between the two reactors. This design enabled demonstration of the technical feasibility of the combination of the circulating fluid-bed reactor system and the iron-based catalyst for commercial hydrogen production. The specific activities associated with this bench-scale circulating fluid-bed reactor systems that were completed in this project included design, construction, commissioning, and operation. The experimental portion of this project focused on technical demonstration of the performance of an iron-based catalyst and a circulating fluid-bed reactor system for hydrogen production. Although a technology can be technically feasible, successful commercial deployment also requires that a technology offer an economic advantage over existing commercial technologies. To effective estimate the economics of this steam-iron process, a techno-economic analysis of this steam iron process and a commercial pressure swing adsorption process were completed. The results from this analysis described in this report show the economic potential of the steam iron process for integration with a gasification plant for coproduction of hydrogen and electricity.« less

Combined cycle plants: Yesterday, today, and tomorrow (review)

NASA Astrophysics Data System (ADS)

Ol'khovskii, G. G.

2016-07-01

Gas turbine plants (GTP) for a long time have been developed by means of increasing the initial gas temperature and improvement of the turbo-machines aerodynamics and the efficiency of the critical components air cooling within the framework of a simple thermodynamic cycle. The application of watercooling systems that were used in experimental turbines and studied approximately 50 years ago revealed the fundamental difficulties that prevented the practical implementation of such systems in the industrial GTPs. The steam cooling researches have developed more substantially. The 300 MW power GTPs with a closedloop steam cooling, connected in parallel with the intermediate steam heating line in the steam cycle of the combined cycle plant (CCP) have been built, tested, and put into operation. The designs and cycle arrangements of such GTPs and entire combined cycle steam plants have become substantially more complicated without significant economic benefits. As a result, the steam cooling of gas turbines has not become widespread. The cycles—complicated by the intermediate air cooling under compression and reheat of the combustion products under expansion and their heat recovery to raise the combustion chamber entry temperature of the air—were used, in particular, in the domestic power GTPs with a moderate (700-800°C) initial gas turbine entry temperature. At the temperatures being reached to date (1300-1450°C), only one company, Alstom, applies in their 240-300 MW GTPs the recycled fuel cycle under expansion of gases in the turbine. Although these GTPs are reliable, there are no significant advantages in terms of their economy. To make a forecast of the further improvement of power GTPs, a brief review and assessment of the water cooling and steam cooling of hot components and complication of the GTP cycle by the recycling of fuel under expansion of gases in the turbine has been made. It is quite likely in the long term to reach the efficiency for the traditional GTPs of approximately 43% and 63% for PGUs at the initial gas temperature of 1600°C and less likely to increase the efficiency of these plants up to 45% and 65% by increasing the gas temperature up to 1700°C or by application of the steam cooling in the recycled fuel cycle.

Investigation and mitigation of condensation induced water hammer by stratified flow experiments

NASA Astrophysics Data System (ADS)

Kadakia, Hiral J.

This research primarily focuses on the possibility of using stratified flow in preventing an occurrence of condensation induced water hammer (CIWH) in horizontal pipe involving steam and subcooled water. A two-phase flow loop simulating the passive safety systems of an advanced light water reactor was constructed and a series of stratified flow experiments were carried out involving a system of subcooled water, saturated water, and steam. Special instruments were designed to measure steam flow rate and subcooled liquid velocity. These experiments showed that when flow field conditions meet certain criteria CIWH does occur. Flow conditions used in experiments were typically observed in passive safety systems of an advanced light water cooled reactor. This research summarizes a) literature research and other experimental data that signify an occurrence of CIWH, b) experiments in an effort to show an occurrence of CIWH and the ability to prevent CIWH, c) qualitative and quantitative results to underline the mechanism of CIWH, d) experiments that show CIWH can be prevented under certain conditions, and e) guidelines for the safe operating conditions. Based on initial experiment results it was observed that Bernoulli's effect can play an important role in wave formation and instability. A separate effect table top experiment was constructed with plexi-glass. A series of entrance effect tests and stratified experiments were carried out with different fluids to study wave formation and wave bridging. Special test series experiments were carried out to investigate the presence of a saturated layer. The effect of subcooled water and steam flow on wedge length and depth were recorded. These experiments helped create a model which calculates wedge and depth of wedge for a given condition of steam and subcooled water. A very good comparison between the experiment results and the model was obtained. These experiments also showed that the presence of saturated layer can mitigate the CIWH. Flow conditions require to mitigate the CIWH must be such that subcooled water is laminar and steam flow rate is less than critical. Finally, a data bank of containing large number of experiments was created and guidelines for safe filling and draining of the system involving steam and subcooled water were created. Also several suggestions are provided to stop CIWH in case it does occur.

Effect of thermal barrier coatings on the performance of steam and water-cooled gas turbine/steam turbine combined cycle system

NASA Technical Reports Server (NTRS)

Nainiger, J. J.

1978-01-01

An analytical study was made of the performance of air, steam, and water-cooled gas-turbine/steam turbine combined-cycle systems with and without thermal-barrier coatings. For steam cooling, thermal barrier coatings permit an increase in the turbine inlet temperature from 1205 C (2200 F), resulting in an efficiency improvement of 1.9 percentage points. The maximum specific power improvement with thermal barriers is 32.4 percent, when the turbine inlet temperature is increased from 1425 C (2600 F) to 1675 C (3050 F) and the airfoil temperature is kept the same. For water cooling, the maximum efficiency improvement is 2.2 percentage points at a turbine inlet temperature of 1683 C (3062 F) and the maximum specific power improvement is 36.6 percent by increasing the turbine inlet temperature from 1425 C (2600 F) to 1730 C (3150 F) and keeping the airfoil temperatures the same. These improvements are greater than that obtained with combined cycles using air cooling at a turbine inlet temperature of 1205 C (2200 F). The large temperature differences across the thermal barriers at these high temperatures, however, indicate that thermal stresses may present obstacles to the use of coatings at high turbine inlet temperatures.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richards, Matt; Hamilton, Chris

This report provides supplemental information to the assessment of target markets provided in Appendix A of the 2012 Next Generation Nuclear Plant (NGNP) Industry Alliance (NIA) business plan [NIA 2012] for deployment of High Temperature Gas-Cooled Reactors (HTGRs) in the 2025 – 2050 time frame. This report largely reiterates the [NIA 2012] assessment for potential deployment of 400 to 800 HTGR modules (100 to 200 HTGR plants with 4 reactor modules) in the 600-MWt class in North America by 2050 for electricity generation, co-generation of steam and electricity, oil sands operations, hydrogen production, and synthetic fuels production (e.g., coal tomore » liquids). As the result of increased natural gas supply from hydraulic fracturing, the current and historically low prices of natural gas remain a significant barrier to deployment of HTGRs and other nuclear reactor concepts in the U.S. However, based on U.S. Department of Energy (DOE) Energy Information Agency (EIA) data, U.S. natural gas prices are expected to increase by the 2030 – 2040 timeframe when a significant number of HTGR modules could be deployed. An evaluation of more recent EIA 2013 data confirms the assumptions in [NIA 2012] of future natural gas prices in the range of approximately $7/MMBtu to $10/MMBtu during the 2030 – 2040 timeframe. Natural gas prices in this range will make HTGR energy prices competitive with natural gas, even in the absence of carbon-emissions penalties. Exhibit ES-1 presents the North American projections in each market segment including a characterization of the market penetration logic. Adjustments made to the 2012 data (and reflected in Exhibit ES-1) include normalization to the slightly larger 625MWt reactor module, segregation between steam cycle and more advanced (higher outlet temperature) modules, and characterization of U.S. synthetic fuel process applications as a separate market segment.« less

Versatile Oxide Films Protect FeCrAl Alloys Under Normal Operation and Accident Conditions in Light Water Power Reactors

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

2018-02-01

The US has currently a fleet of 99 nuclear power light water reactors which generate approximately 20% of the electricity consumed in the country. Near 90% of the reactors are at least 30 years old. There are incentives to make the existing reactors safer by using accident tolerant fuels (ATF). Compared to the standard UO2-zirconium-based system, ATF need to tolerate loss of active cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. Ferritic iron-chromium-aluminum (FeCrAl) alloys have been identified as an alternative to replace current zirconium alloys. They contain Fe (base) + 10-22 Cr + 4-6 Al and may contain smaller amounts of other elements such as molybdenum and traces of others. FeCrAl alloys offer outstanding resistance to attack by superheated steam by developing an alumina oxide on the surface in case of a loss of coolant accident like at Fukushima. FeCrAl alloys also perform well under normal operation conditions both in boiling water reactors and pressurized water reactors because they are protected by a thin oxide rich in chromium. Under normal operation condition, the key element is Cr and under accident conditions it is Al.

Exit chimney joint and method of forming the joint for closed circuit steam cooled gas turbine nozzles

DOEpatents

Burdgick, Steven Sebastian; Burns, James Lee

2002-01-01

A nozzle segment for a gas turbine includes inner and outer band portions and a vane extending between the band portions. The inner and outer band portions are each divided into first and second plenums separated by an impingement plate. Cooling steam is supplied to the first cavity for flow through the apertures to cool the outer nozzle wall. The steam flows through a leading edge cavity in the vane into the first cavity of the inner band portion for flow through apertures of the impingement plate to cool the inner nozzle wall. Spent cooling steam flows through a plurality of cavities in the vane, exiting through an exit chimney in the outer band. The exit chimney is secured at its inner end directly to the nozzle vane wall surrounding the exit cavities, to the margin of the impingement plate at a location intermediate the ends of the exit chimney and to margins of an opening through the cover whereby each joint is externally accessible for joint formation and for subsequent inspection.

Preliminary assessment of alternative PFBC power plant systems. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wysocki, J.; Rogali, R.

1980-07-01

This report presents the design and and economic comparisons of the following nominal 1000 MWe PFBC power plants for both eastern and western coal: Curtiss-Wright PFBC power plants with an air-cooled design; General Electric RFBC power plants with a steam-cooled design; and AEP/Stal-Laval PFBC power plants with a steam-cooled design. In addition, reference pulverized coal-fired (PCF) power plants are included for comparison purposes. The results of the analysis indicate: (1) The steam-cooled PFBC designs show potential savings of 10% and 11% over PCF plants for eastern and western coal, respectively, in terms of busbar power cost; (2) the air-cooled PFBCmore » designs show potential savings of 1% and 2% over PCF plants for eastern and western coal, respectively, in terms of busbar power cost.« less

Practical achievements on biomass steam gasification in a rotary tubular coiled-downdraft reactor.

PubMed

Andrew, Renny; Gokak, D T; Sharma, Pankaj; Gupta, Shalini

2016-12-01

Today, the impending stringent environmental norms and concerns about the depletion of fossil fuel reserves have added impetus on development of cutting edge technologies for production of alternative fuels from renewable sources, like biomass. The concept of biomass pyro-gasification offers a platform for production of (a) hydrogen, (b) hydrocarbons and (c) value added chemicals, etc. In this context, there exists potential for hydrogen production from biomass by superheated steam gasification. Apart from H 2 , gaseous products of biomass steam gasification contain CO, CH 4 and other hydrocarbons that can be converted to hydrogen through cracking, steam reforming and water gas shift reactions. In the present work, the characteristics of biomass steam gasification in an indigenously designed rotary tubular coiled-downdraft reactor for high value gaseous fuel production from rice husk was studied through a series of experiments. The robust reactor system enhances biomass conversion to gaseous products by improved mass and heat transfer within the system induced by a coiled flow pattern with increased heat transfer area. Also, the system has improved upon the reliability of operation and offered greater continuity of the process and easier control in comparison with a conventional process by making use of an innovative gas cooler assembly and efficient venturi-mixing system for biomass and steam. Subsequently, the effects of reactor temperature, steam-to-biomass ratio and residence time on overall product gas yield and hydrogen yield were investigated. From the experimental results, it can be deduced that an optimum reactor temperature of 750 °C, steam-to-biomass ratio of 2.0 and a residence time of 3.0 min contributed highest gas yield (1.252 Nm 3  kg -1 moisture-free biomass). Based on the obtained experimental results, a projected potential hydrogen yield of 8.6 wt% of the moisture-free biomass could be achieved, and is also practical for production of pure hydrogen. © The Author(s) 2016.

Steam conversion of liquefied petroleum gas and methane in microchannel reactor

NASA Astrophysics Data System (ADS)

Dimov, S. V.; Gasenko, O. A.; Fokin, M. I.; Kuznetsov, V. V.

2018-03-01

This study presents experimental results of steam conversion of liquefied petroleum gas and methane in annular catalytic reactor - heat exchanger. The steam reforming was done on the Rh/Al2O3 nanocatalyst with the heat applied through the microchannel gap from the outer wall. Concentrations of the products of chemical reactions in the outlet gas mixture are measured at different temperatures of reactor. The range of channel wall temperatures at which the ratio of hydrogen and carbon oxide in the outlet mixture grows substantially is determined. Data on the composition of liquefied petroleum gas conversion products for the ratio S/C = 5 was received for different GHVS.

PANDA asymmetric-configuration passive decay heat removal test results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fischer, O.; Dreier, J.; Aubert, C.

1997-12-01

PANDA is a large-scale, low-pressure test facility for investigating passive decay heat removal systems for the next generation of LWRs. In the first series of experiments, PANDA was used to examine the long-term LOCA response of the Passive Containment Cooling System (PCCS) for the General Electric (GE) Simplified Boiling Water Reactor (SBWR). The test objectives include concept demonstration and extension of the database available for qualification of containment codes. Also included is the study of the effects of nonuniform distributions of steam and noncondensable gases in the Dry-well (DW) and in the Suppression Chamber (SC). 3 refs., 9 figs.

Alloy Selection for Accident Tolerant Fuel Cladding in Commercial Light Water Reactors

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

2015-12-01

As a consequence of the March 2011 events at the Fukushima site, the U.S. congress asked the Department of Energy (DOE) to concentrate efforts on the development of nuclear fuels with enhanced accident tolerance. The new fuels had to maintain or improve the performance of current UO2-zirconium alloy rods during normal operation conditions and tolerate the loss of active cooling in the core for a considerably longer time period than the current system. DOE is funding cost-shared research to investigate the behavior of advanced steels both under normal operation conditions in high-temperature water [ e.g., 561 K (288 °C)] and under accident conditions for reaction with superheated steam. Current results show that, under accident conditions, the advanced ferritic steels (1) have orders of magnitude lower reactivity with steam, (2) would generate less hydrogen and heat than the current zirconium alloys, (3) are resistant to stress corrosion cracking under normal operation conditions, and (4) have low general corrosion in water at 561 K (288 °C).

French Regulatory practice and experience feedback on steam generator tube integrity

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sandon, G.

1997-02-01

This paper summarizes the way the French Safety Authority applies regulatory rules and practices to the problem of steam generator tube cracking in French PWR reactors. There are 54 reactors providing 80% of French electrical consumption. The Safety Authority closely monitors the performance of tubes in steam generators, and requires application of a program which deals with problems prior to the actual development of leakage. The actual rules regarding such performance are flexible, responding to the overall performance of operating steam generators. In addition there is an inservice inspection service to examine tubes during shutdown, and to monitor steam generatorsmore » for leakage during operation, with guidelines for when generators must be pulled off line.« less

Determination of the steam volume fraction in the event of loss of cooling of the spent fuel storage pool

NASA Astrophysics Data System (ADS)

Sledkov, R. M.; Galkin, I. Yu.; Stepanov, O. E.; Strebnev, N. A.

2017-01-01

When one solves engineering problems related to the cooling of fuel assemblies (FAs) in a spent fuel storage pool (SFSP) and the assessment of nuclear safety of FA storage in an SFSP in the initial event of loss of SFSP cooling, it is essential to determine the coolant density and, consequently, steam volume fractions φ in bundles of fuel elements at a pressure of 0.1-0.5 MPa. Such formulas for calculating φ that remain valid in a wide range of operating parameters and geometric shapes of channels and take the conditions of loss of SFSP cooling into account are currently almost lacking. The results of systematization and analysis of the available formulas for φ are reported in the present study. The calculated values were compared with the experimental data obtained in the process of simulating the conditions of FA cooling in an SFSP in the event of loss of its cooling. Six formulas for calculating the steam volume fraction, which were used in this comparison, were chosen from a total of 11 considered relations. As a result, the formulas producing the most accurate values of φ in the conditions of loss of SFSP cooling were selected. In addition, a relation that allows one to perform more accurate calculations of steam volume fractions in the conditions of loss of SFSP cooling was derived based on the Fedorov formula in the two-group approximation.

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.

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.

Method to prevent/mitigate steam explosions in casting pits

DOEpatents

Taleyarkhan, Rusi P.

1996-01-01

Steam explosions can be prevented or mitigated during a metal casting process by the placement of a perforated flooring system in the casting pit. An upward flow of compressed gas through this perforated flooring system is introduced during the casting process to produce a buffer layer between any spilled molten metal and the cooling water in the reservoir. This buffer layer provides a hydrodynamic layer which acts to prevent or mitigate steam explosions resulting from hot, molten metal being spilled into or onto the cooling water.

Study on Material Selection of Reactor Pressure Vessel of SCWR

NASA Astrophysics Data System (ADS)

Ma, Shuli; Luo, Ying; Yin, Qinwei; Li, Changxiang; Xie, Guofu

This paper first analyzes the feasibility of SA-508 Grade 3 Class 1 Steel as an alternative material for Supercritical Water-Cooled Reactor (SCWR) Reactor Pressure Vessel (RPV). This kind of steel is limited to be applied in SCWR RPV due to its quenching property, though large forging could be accomplished by domestic manufacturers in forging aspect. Therefore, steels with higher strength and better quenching property are needed for SWCR RPV. The chemical component of SA-508 Gr.3 Cl.2 steel is similar to that of SA-508 Gr.3 Cl.1 steel, and more appropriate matching of strength and toughness could be achieved by the adjusting the elements contents, as well as proper control of tempering temperature and time. In light of the fact that Cl.2 steel has been successfully applied to steam generator, it could be an alternative material for SWCR RPV. SA-508 Gr.4N steel with high strength and good toughness is another alternative material for SCWR RPV. But large amount of research work before application is still needed for the lack of data on welding and irradiation etc.

Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

NASA Astrophysics Data System (ADS)

Tyapkov, V. F.

2014-07-01

The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

BOILING SLURRY REACTOR AND METHOD FO CONTROL

DOEpatents

Petrick, M.; Marchaterre, J.F.

1963-05-01

The control of a boiling slurry nuclear reactor is described. The reactor consists of a vertical tube having an enlarged portion, a steam drum at the top of the vertical tube, and at least one downcomer connecting the steam drum and the bottom of the vertical tube, the reactor being filled with a slurry of fissionabie material in water of such concentration that the enlarged portion of the vertical tube contains a critical mass. The slurry boils in the vertical tube and circulates upwardly therein and downwardly in the downcomer. To control the reactor by controlling the circulation of the slurry, a gas is introduced into the downcomer. (AEC)

Chemical compatibility issues associated with use of SiC/SiC in advanced reactor concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilson, Dane F.

2015-09-01

Silicon carbide/silicon carbide (SiC/SiC) composites are of interest for components that will experience high radiation fields in the High Temperature Gas Cooled Reactor (HTGR), the Very High Temperature Reactor (VHTR), the Sodium Fast Reactor (SFR), or the Fluoride-cooled High-temperature Reactor (FHR). In all of the reactor systems considered, reactions of SiC/SiC composites with the constituents of the coolant determine suitability of materials of construction. The material of interest is nuclear grade SiC/SiC composites, which consist of a SiC matrix [high-purity, chemical vapor deposition (CVD) SiC or liquid phase-sintered SiC that is crystalline beta-phase SiC containing small amounts of alumina-yttria impurity],more » a pyrolytic carbon interphase, and somewhat impure yet crystalline beta-phase SiC fibers. The interphase and fiber components may or may not be exposed, at least initially, to the reactor coolant. The chemical compatibility of SiC/SiC composites in the three reactor environments is highly dependent on thermodynamic stability with the pure coolant, and on reactions with impurities present in the environment including any ingress of oxygen and moisture. In general, there is a dearth of information on the performance of SiC in these environments. While there is little to no excess Si present in the new SiC/SiC composites, the reaction of Si with O 2 cannot be ignored, especially for the FHR, in which environment the product, SiO 2, can be readily removed by the fluoride salt. In all systems, reaction of the carbon interphase layer with oxygen is possible especially under abnormal conditions such as loss of coolant (resulting in increased temperature), and air and/ or steam ingress. A global outline of an approach to resolving SiC/SiC chemical compatibility concerns with the environments of the three reactors is presented along with ideas to quickly determine the baseline compatibility performance of SiC/SiC.« less

Component testing of a ground based gas turbine steam cooled rich-burn primary zone combustor for emissions control of nitrogeneous fuels

NASA Technical Reports Server (NTRS)

Schultz, D. F.

1986-01-01

This effort summarizes the work performed on a steam cooled, rich-burn primary zone, variable geometry combustor designed for combustion of nitrogeneous fuels such as heavy oils or synthetic crude oils. The steam cooling was employed to determine its feasibility and assess its usefulness as part of a ground based gas turbine bottoming cycle. Variable combustor geometry was employed to demonstrate its ability to control primary and secondary zone equivalence ratios and overall pressure drop. Both concepts proved to be highly successful in achieving their desired objectives. The steam cooling reduced peak liner temperatures to less than 800 K. This low temperature offers the potential of both long life and reduced use of strategic materials for liner fabrication. These degrees of variable geometry were successfully employed to control air flow distribution within the combustor. A variable blade angle axial flow air swirler was used to control primary zone air flow, while the secondary and tertiary zone air flows were controlled by rotating bands which regulated air flow to the secondary zone quench holes and the dilutions holes respectively.

Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

DOEpatents

Hunsbedt, Anstein; Busboom, Herbert J.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

Heat and fuel coupled operation of a high temperature polymer electrolyte fuel cell with a heat exchanger methanol steam reformer

NASA Astrophysics Data System (ADS)

Schuller, G.; Vázquez, F. Vidal; Waiblinger, W.; Auvinen, S.; Ribeirinha, P.

2017-04-01

In this work a methanol steam reforming (MSR) reactor has been operated thermally coupled to a high temperature polymer electrolyte fuel cell stack (HT-PEMFC) utilizing its waste heat. The operating temperature of the coupled system was 180 °C which is significantly lower than the conventional operating temperature of the MSR process which is around 250 °C. A newly designed heat exchanger reformer has been developed by VTT (Technical Research Center of Finland LTD) and was equipped with commercially available CuO/ZnO/Al2O3 (BASF RP-60) catalyst. The liquid cooled, 165 cm2, 12-cell stack used for the measurements was supplied by Serenergy A/S. The off-heat from the electrochemical fuel cell reaction was transferred to the reforming reactor using triethylene glycol (TEG) as heat transfer fluid. The system was operated up to 0.4 A cm-2 generating an electrical power output of 427 Wel. A total stack waste heat utilization of 86.4% was achieved. It has been shown that it is possible to transfer sufficient heat from the fuel cell stack to the liquid circuit in order to provide the needed amount for vaporizing and reforming of the methanol-water-mixture. Furthermore a set of recommendations is given for future system design considerations.

Startup of air-cooled condensers and dry cooling towers at low temperatures of the cooling air

NASA Astrophysics Data System (ADS)

Milman, O. O.; Ptakhin, A. V.; Kondratev, A. V.; Shifrin, B. A.; Yankov, G. G.

2016-05-01

The problems of startup and performance of air-cooled condensers (ACC) and dry cooling towers (DCT) at low cooling air temperatures are considered. Effects of the startup of the ACC at sub-zero temperatures are described. Different options of the ACC heating up are analyzed, and examples of existing technologies are presented (electric heating, heating up with hot air or steam, and internal and external heating). The use of additional heat exchanging sections, steam tracers, in the DCT design is described. The need for high power in cases of electric heating and heating up with hot air is noted. An experimental stand for research and testing of the ACC startup at low temperatures is described. The design of the three-pass ACC unit is given, and its advantages over classical single-pass design at low temperatures are listed. The formation of ice plugs inside the heat exchanging tubes during the start-up of ACC and DCT at low cooling air temperatures is analyzed. Experimental data on the effect of the steam flow rate, steam nozzle distance from the heat-exchange surface, and their orientation in space on the metal temperature were collected, and test results are analyzed. It is noted that the surface temperature at the end of the heat up is almost independent from its initial temperature. Recommendations for the safe start-up of ACCs and DCTs are given. The heating flow necessary to sufficiently heat up heat-exchange surfaces of ACCs and DCTs for the safe startup is estimated. The technology and the process of the heat up of the ACC with the heating steam external supply are described by the example of the startup of the full-scale section of the ACC at sub-zero temperatures of the cooling air, and the advantages of the proposed start-up technology are confirmed.

Passive cooling safety system for liquid metal cooled nuclear reactors

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Indirect passive cooling system for liquid metal cooled nuclear reactors

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1990-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Apparatus and methods for impingement cooling of an undercut region adjacent a side wall of a turbine nozzle segment

DOEpatents

Burdgick, Steven Sebastian; Itzel, Gary Michael

2001-01-01

A gas turbine nozzle segment has outer and inner bands. Each band includes a side wall, a cover and an impingement plate between the cover and nozzle wall defining two cavities on opposite sides of the impingement plate. Cooling steam is supplied to one cavity for flow through apertures of the impingement plate to cool the nozzle wall. The side wall of the band and inturned flange define with the nozzle wall an undercut region. The inturned flange has a plurality of apertures for directing cooling steam to cool the side wall between adjacent nozzle segments.

Solar coal gasification reactor with pyrolysis gas recycle

DOEpatents

Aiman, William R.; Gregg, David W.

1983-01-01

Coal (or other carbonaceous matter, such as biomass) is converted into a duct gas that is substantially free from hydrocarbons. The coal is fed into a solar reactor (10), and solar energy (20) is directed into the reactor onto coal char, creating a gasification front (16) and a pyrolysis front (12). A gasification zone (32) is produced well above the coal level within the reactor. A pyrolysis zone (34) is produced immediately above the coal level. Steam (18), injected into the reactor adjacent to the gasification zone (32), reacts with char to generate product gases. Solar energy supplies the energy for the endothermic steam-char reaction. The hot product gases (38) flow from the gasification zone (32) to the pyrolysis zone (34) to generate hot char. Gases (38) are withdrawn from the pyrolysis zone (34) and reinjected into the region of the reactor adjacent the gasification zone (32). This eliminates hydrocarbons in the gas by steam reformation on the hot char. The product gas (14) is withdrawn from a region of the reactor between the gasification zone (32) and the pyrolysis zone (34). The product gas will be free of tar and other hydrocarbons, and thus be suitable for use in many processes.

High aspect ratio catalytic reactor and catalyst inserts therefor

DOEpatents

Lin, Jiefeng; Kelly, Sean M.

2018-04-10

The present invention relates to high efficient tubular catalytic steam reforming reactor configured from about 0.2 inch to about 2 inch inside diameter high temperature metal alloy tube or pipe and loaded with a plurality of rolled catalyst inserts comprising metallic monoliths. The catalyst insert substrate is formed from a single metal foil without a central supporting structure in the form of a spiral monolith. The single metal foil is treated to have 3-dimensional surface features that provide mechanical support and establish open gas channels between each of the rolled layers. This unique geometry accelerates gas mixing and heat transfer and provides a high catalytic active surface area. The small diameter, high aspect ratio tubular catalytic steam reforming reactors loaded with rolled catalyst inserts can be arranged in a multi-pass non-vertical parallel configuration thermally coupled with a heat source to carry out steam reforming of hydrocarbon-containing feeds. The rolled catalyst inserts are self-supported on the reactor wall and enable efficient heat transfer from the reactor wall to the reactor interior, and lower pressure drop than known particulate catalysts. The heat source can be oxygen transport membrane reactors.

Development of coal-feeding systems at the Morgantown Energy Research Center

NASA Technical Reports Server (NTRS)

Hobday, J. M.

1977-01-01

Systems for feeding crushed and pulverized coal into coal conversion reactor vessels are described. Pneumatic methods for feeding pulverized coal, slurry feeders, and coal pumps, methods for steam pickup, and a method for drying a water-coal slurry in a steam fluidized bed subsequent to feeding the coal into a reactor vessel are included.

Concentrating Solar Power Projects - Olivenza 1 | Concentrating Solar Power

Science.gov Websites

Manufacturer: Siemens Turbine Description: 5 extractions Output Type: Steam Rankine Power Cycle Pressure: 100.0 bar Cooling Method: Wet cooling Cooling Method Description: Cooling Towers

One-Step Coal Liquefaction

NASA Technical Reports Server (NTRS)

Qader, S. A.

1984-01-01

Steam injection improves yield and quality of product. Single step process for liquefying coal increases liquid yield and reduces hydrogen consumption. Principal difference between this and earlier processes includes injection of steam into reactor. Steam lowers viscosity of liquid product, so further upgrading unnecessary.

Method to prevent/mitigate steam explosions in casting pits

DOEpatents

Taleyarkhan, R.P.

1996-12-24

Steam explosions can be prevented or mitigated during a metal casting process by the placement of a perforated flooring system in the casting pit. An upward flow of compressed gas through this perforated flooring system is introduced during the casting process to produce a buffer layer between any spilled molten metal and the cooling water in the reservoir. This buffer layer provides a hydrodynamic layer which acts to prevent or mitigate steam explosions resulting from hot, molten metal being spilled into or onto the cooling water. 3 figs.

APPARATUS FOR CONTROL OF A BOILING REACTOR RESPONSIVE TO STEAM DEMAND

DOEpatents

Treshow, M.

1963-07-23

A method of controlling a fuel-rod-in-tube-type boilingwater reactor having nozzles at the point of water entry into the tube is described. Water is pumped into the nozzles by an auxiliary pump operated by steam from an interstage position of the associated turbine, so that the pumping speed is responsive to turbine demand. (AEC)

Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1993-01-01

A liquid metal cooled nuclear fission reactor plant having a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during reactor shutdown, or heat produced during a mishap. This reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary system when rendered inoperable.

Dry coolers and air-condensing units (Review)

NASA Astrophysics Data System (ADS)

Milman, O. O.; Anan'ev, P. A.

2016-03-01

The analysis of factors affecting the growth of shortage of freshwater is performed. The state and dynamics of the global market of dry coolers used at electric power plants are investigated. Substantial increase in number and maximum capacity of air-cooled condensers, which have been put into operation in the world in recent years, are noted. The key reasons facilitating the choice of developers of the dry coolers, in particular the independence of the location of thermal power plant from water sources, are enumerated. The main steam turbine heat removal schemes using air cooling are considered, their comparison of thermal efficiency is assessed, and the change of three important parameters, such as surface area of heat transfer, condensate pump flow, and pressure losses in the steam exhaust system, are estimated. It is shown that the most effective is the scheme of direct steam condensation in the heat-exchange tubes, but other schemes also have certain advantages. The air-cooling efficiency may be enhanced much more by using an air-cooling hybrid system: a combination of dry and wet cooling. The basic applied constructive solutions are shown: the arrangement of heat-exchange modules and the types of fans. The optimal mounting design of a fully shopassembled cooling system for heat-exchange modules is represented. Different types of heat-exchange tubes ribbing that take into account the operational features of cooling systems are shown. Heat transfer coefficients of the plants from different manufacturers are compared, and the main reasons for its decline are named. When using evaporative air cooling, it is possible to improve the efficiency of air-cooling units. The factors affecting the faultless performance of dry coolers (DC) and air-condensing units (ACU) and the ways of their elimination are described. A high velocity wind forcing reduces the efficiency of cooling systems and creates preconditions for the development of wind-driven devices. It is noted that global trends have a significant influence on the application of dry coolers in Russia, in view of the fact that some TPP have a surface condensers arrangement. The reasons that these systems are currently less efficient than the direct steam condensation in an air-cooled condenser are explained. It is shown that, in some cases, it is more reasonable to use mixing-type condensers in combination with a dry cooler. Measures for a full import substitution of steam exhaust heat removal systems are mentioned.

Process assessment of small scale low temperature methanol synthesis

NASA Astrophysics Data System (ADS)

Hendriyana, Susanto, Herri; Subagjo

2015-12-01

Biomass is a renewable energy resource and has the potential to make a significant impact on domestic fuel supplies. Biomass can be converted to fuel like methanol via several step process. The process can be split into following main steps: biomass preparation, gasification, gas cooling and cleaning, gas shift and methanol synthesis. Untill now these configuration still has a problem like high production cost, catalyst deactivation, economy of scale and a huge energy requirements. These problems become the leading inhibition for biomass conversion to methanol, which should be resolved to move towards the economical. To address these issues, we developed various process and new configurations for methanol synthesis via methyl formate. This configuration combining two reactors: the one reactor for the carbonylation of methanol and CO to form methyl formate, and the second for the hydrogenolysis of methyl formate and H2 to form two molecule of methanol. Four plant process configurations were compared with the biomass basis is 300 ton/day. The first configuration (A) is equipped with a steam reforming process for converting methane to CO and H2 for increasing H2/CO ratio. CO2 removal is necessary to avoid poisoning the catalyst. COSORB process used for the purpose of increasing the partial pressure of CO in the feed gas. The steam reforming process in B configuration is not used with the aim of reducing the number of process equipment, so expect lower investment costs. For C configuration, the steam reforming process and COSORB are not used with the aim of reducing the number of process equipment, so expect lower investment costs. D configuration is almost similar to the configuration A. This configuration difference is in the synthesis of methanol which was held in a single reactor. Carbonylation and hydrogenolysis reactions carried out in the same reactor one. These processes were analyzed in term of technical process, material and energy balance and economic analysis. The presented study is an attempt to compile most of these efforts in order to guide future work to get cheaper low cost investment. From our study the interesting configuration to the next development is D configuration with methanol yield 112 ton/day and capital cost with 526.4 106 IDR. The configuration of D with non-discounted and discounted rate had the break-even point approximately six and eight years.

Thermal Propulsion Capture System Heat Exchanger Design

NASA Technical Reports Server (NTRS)

Richard, Evan M.

2016-01-01

One of the biggest challenges of manned spaceflight beyond low earth orbit and the moon is harmful radiation that astronauts would be exposed to on their long journey to Mars and further destinations. Using nuclear energy has the potential to be a more effective means of propulsion compared to traditional chemical engines (higher specific impulse). An upper stage nuclear engine would allow astronauts to reach their destination faster and more fuel efficiently. Testing these engines poses engineering challenges due to the need to totally capture the engine exhaust. The Thermal Propulsion Capture System is a concept for cost effectively and safely testing Nuclear Thermal Engines. Nominally, hydrogen exhausted from the engine is not radioactive, but is treated as such in case of fuel element failure. The Thermal Propulsion Capture System involves injecting liquid oxygen to convert the hydrogen exhaust into steam. The steam is then cooled and condensed into liquid water to allow for storage. The Thermal Propulsion Capture System concept for ground testing of a nuclear powered engine involves capturing the engine exhaust to be cooled and condensed before being stored. The hydrogen exhaust is injected with liquid oxygen and burned to form steam. That steam must be cooled to saturation temperatures before being condensed into liquid water. A crossflow heat exchanger using water as a working fluid will be designed to accomplish this goal. Design a cross flow heat exchanger for the Thermal Propulsion Capture System testing which: Eliminates the need for water injection cooling, Cools steam from 5800 F to saturation temperature, and Is efficient and minimizes water requirement.

UTSG-2; A theoretical model describing the transient behavior of a pressurized water reactor natural circulation U-tube steam generator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hold, A.

An advanced nonlinear transient model for calculating steady-state and dynamic behaviors of characteristic parameters of a Kraftwerk Union-type vertical natural-circulation U-tube steam generator and its main steam system is presented. This model has been expanded due to the increasing need for safety-related accident research studies. It now takes into consideration the possibilities of dryout and superheating along the secondary side of the steam generator. The resulting theoretical model is the basis of the digital code UTSG-2, which can be used both by itself and in combination with other pressurized water reactor transient codes, such as ALMOD-3.4, AMOD-4, and ATHLET.

Central waste processing system

NASA Technical Reports Server (NTRS)

Kester, F. L.

1973-01-01

A new concept for processing spacecraft type wastes has been evaluated. The feasibility of reacting various waste materials with steam at temperatures of 538 - 760 C in both a continuous and batch reactor with residence times from 3 to 60 seconds has been established. Essentially complete gasification is achieved. Product gases are primarily hydrogen, carbon dioxide, methane, and carbon monoxide. Water soluble synthetic wastes are readily processed in a continuous tubular reactor at concentrations up to 20 weight percent. The batch reactor is able to process wet and dry wastes at steam to waste weight ratios from 2 to 20. Feces, urine, and synthetic wastes have been successfully processed in the batch reactor.

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fanning, T. H.; Brunett, A. J.; Sumner, T.

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transientmore » thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.« less

Liquid metal cooled nuclear reactors with passive cooling system

DOEpatents

Hunsbedt, Anstein; Fanning, Alan W.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

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.

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.

BWR Steam Dryer Alternating Stress Assessment Procedures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morante, R. J.; Hambric, S. A.; Ziada, S.

2016-12-01

This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

Pressure intelligent control strategy of Waste heat recovery system of converter vapors

NASA Astrophysics Data System (ADS)

Feng, Xugang; Wu, Zhiwei; Zhang, Jiayan; Qian, Hong

2013-01-01

The converter gas evaporative cooling system is mainly used for absorbing heat in the high temperature exhaust gas which produced by the oxygen blowing reaction. Vaporization cooling steam pressure control system of converter is a nonlinear, time-varying, lagging behind, close coupling of multivariable control object. This article based on the analysis of converter operation characteristics of evaporation cooling system, of vaporization in a production run of pipe pressure variation and disturbance factors.For the dynamic characteristics of the controlled objects,we have improved the conventional PID control scheme.In Oxygen blowing process, we make intelligent control by using fuzzy-PID cascade control method and adjusting the Lance,that it can realize the optimization of the boiler steam pressure control.By design simulation, results show that the design has a good control not only ensures drum steam pressure in the context of security, enabling efficient conversion of waste heat.And the converter of 1800 flue gas through pipes and cool and dust removal also can be cooled to about 800. Therefore the converter haze evaporative cooling system has achieved to the converter haze temperature decrease effect and enhanced to the coal gas returns-ratio.

Nanoparticulate-catalyzed oxygen transfer processes

DOEpatents

Hunt, Andrew T [Atlanta, GA; Breitkopf, Richard C [Dunwoody, GA

2009-12-01

Nanoparticulates of oxygen transfer materials that are oxides of rare earth metals, combinations of rare earth metals, and combinations of transition metals and rare earth metals are used as catalysts in a variety of processes. Unexpectedly large thermal efficiencies are achieved relative to micron sized particulates. Processes that use these catalysts are exemplified in a multistage reactor. The exemplified reactor cracks C6 to C20 hydrocarbons, desulfurizes the hydrocarbon stream and reforms the hydrocarbons in the stream to produce hydrogen. In a first reactor stage the steam and hydrocarbon are passed through particulate mixed rare earth metal oxide to crack larger hydrocarbon molecules. In a second stage, the steam and hydrocarbon are passed through particulate material that desulfurizes the hydrocarbon. In a third stage, the hydrocarbon and steam are passed through a heated, mixed transition metal/rare earth metal oxide to reform the lower hydrocarbons and thereby produce hydrogen. Stages can be alone or combined. Parallel reactors can provide continuous reactant flow. Each of the processes can be carried out individually.

Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pahari, S.; Hajela, S.; Rammohan, H. P.

2012-07-01

700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG andmore » PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Brien, James E.; Sabharwall, Piyush; Yoon, Su -Jong

2014-09-01

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

Thermal Aspects of Using Alternative Nuclear Fuels in Supercritical Water-Cooled Reactors

NASA Astrophysics Data System (ADS)

Grande, Lisa Christine

A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT)-type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.

Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

NASA Astrophysics Data System (ADS)

Wright, Steven A.; Houts, Michael

2001-02-01

Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

System and method for determining coolant level and flow velocity in a nuclear reactor

DOEpatents

Brisson, Bruce William; Morris, William Guy; Zheng, Danian; Monk, David James; Fang, Biao; Surman, Cheryl Margaret; Anderson, David Deloyd

2013-09-10

A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

POWER GENERATING NEUTRONIC REACTOR SYSTEM

DOEpatents

Vernon, H.C.

1958-03-01

This patent relates to reactor systems of the type wherein the cooiing medium is a liquid which is converted by the heat of the reaction to steam which is conveyed directly to a pnime mover such as a steam turbine driving a generatore after which it is condensed and returred to the coolant circuit. In this design, the reactor core is disposed within a tank for containing either a slurry type fuel or an aggregation of solid fuel elements such as elongated rods submerged in a liquid moderator such as heavy water. The top of the tank is provided with a nozzle which extends into an expansion chamber connected with the upper end of the tank, the coolant being maintained in the expansion chamber at a level above the nozzle and the steam being formed in the expansion chamber.

Solvent refined coal reactor quench system

DOEpatents

Thorogood, Robert M.

1983-01-01

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

Solvent refined coal reactor quench system

DOEpatents

Thorogood, R.M.

1983-11-08

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

NEUTRONIC REACTOR SYSTEM

DOEpatents

Treshow, M.

1959-02-10

A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

Neutron fluxes in test reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youinou, Gilles Jean-Michel

Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

Ceramic oxygen transport membrane array reactor and reforming method

DOEpatents

Kelly, Sean M.; Christie, Gervase Maxwell; Robinson, Charles; Wilson, Jamie R; Gonzalez, Javier E.; Doraswami, Uttam R.

2017-10-03

The invention relates to a commercially viable modular ceramic oxygen transport membrane system for utilizing heat generated in reactively-driven oxygen transport membrane tubes to generate steam, heat process fluid and/or provide energy to carry out endothermic chemical reactions. The system provides for improved thermal coupling of oxygen transport membrane tubes to steam generation tubes or process heater tubes or reactor tubes for efficient and effective radiant heat transfer.

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.

Lubricating system for thermal medium delivery parts in a gas turbine

DOEpatents

Mashey, Thomas Charles

2002-01-01

Cooling steam delivery tubes extend axially along the outer rim of a gas turbine rotor for supplying cooling steam to and returning spent cooling steam from the turbine buckets. Because of the high friction forces at the interface of the tubes and supporting elements due to rotor rotation, a low coefficient of friction coating is provided at the interface of the tubes and support elements. On each surface, a first coating of a cobalt-based alloy is sprayed onto the surface at high temperature. A portion of the first coating is machined off to provide a smooth, hard surface. A second ceramic-based solid film lubricant is sprayed onto the first coating. By reducing the resistance to axial displacement of the tubes relative to the supporting elements due to thermal expansion, the service life of the tubes is substantially extended.

Screening reactor steam/water piping systems for water hammer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Griffith, P.

1997-09-01

A steam/water system possessing a certain combination of thermal, hydraulic and operational states, can, in certain geometries, lead to a steam bubble collapse induced water hammer. These states, operations, and geometries are identified. A procedure that can be used for identifying whether an unbuilt reactor system is prone to water hammer is proposed. For the most common water hammer, steam bubble collapse induced water hammer, six conditions must be met in order for one to occur. These are: (1) the pipe must be almost horizontal; (2) the subcooling must be greater than 20 C; (3) the L/D must be greatermore » than 24; (4) the velocity must be low enough so that the pipe does not run full, i.e., the Froude number must be less than one; (5) there should be void nearby; (6) the pressure must be high enough so that significant damage occurs, that is the pressure should be above 10 atmospheres. Recommendations on how to avoid this kind of water hammer in both the design and the operation of the reactor system are made.« less

Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 1: Laboratory Experiments and Application to EBR-II Secondary Sodium System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steven R. Sherman

Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/ormore » to comply with decommissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidified carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, U.S.A. This report is Part 1 of a two-part report. It is divided into three sections. The first section describes the chemistry of carbon dioxide-water-sodium reactions. The second section covers the laboratory experiments that were conducted in order to develop the residual sodium deactivation process. The third section discusses the application of the deactivation process to the treatment of residual sodium within the EBR-II secondary sodium cooling system. Part 2 of the report, under separate cover, describes the application of the technique to residual sodium treatment within the EBR-II primary sodium cooling system and related systems.« less

43 CFR 3276.13 - What additional information must I give BLM in the monthly report for flash and dry steam...

Code of Federal Regulations, 2014 CFR

2014-10-01

... flow into the turbine in klbs; for dual flash facilities, you must separate the steam flow into high...; (d) Auxiliary steam flow used for gas ejectors, steam seals, pumps, etc., in klbs; (e) Flow of condensate out of the plant (after the cooling towers) in klbs; and (f) Any other information we may require. ...

43 CFR 3276.13 - What additional information must I give BLM in the monthly report for flash and dry steam...

Code of Federal Regulations, 2012 CFR

2012-10-01

... flow into the turbine in klbs; for dual flash facilities, you must separate the steam flow into high...; (d) Auxiliary steam flow used for gas ejectors, steam seals, pumps, etc., in klbs; (e) Flow of condensate out of the plant (after the cooling towers) in klbs; and (f) Any other information we may require. ...

43 CFR 3276.13 - What additional information must I give BLM in the monthly report for flash and dry steam...

Code of Federal Regulations, 2011 CFR

2011-10-01

... flow into the turbine in klbs; for dual flash facilities, you must separate the steam flow into high...; (d) Auxiliary steam flow used for gas ejectors, steam seals, pumps, etc., in klbs; (e) Flow of condensate out of the plant (after the cooling towers) in klbs; and (f) Any other information we may require. ...

43 CFR 3276.13 - What additional information must I give BLM in the monthly report for flash and dry steam...

Code of Federal Regulations, 2013 CFR

2013-10-01

... flow into the turbine in klbs; for dual flash facilities, you must separate the steam flow into high...; (d) Auxiliary steam flow used for gas ejectors, steam seals, pumps, etc., in klbs; (e) Flow of condensate out of the plant (after the cooling towers) in klbs; and (f) Any other information we may require. ...

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 1: Objectives, summary results and introduction

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

The objective was to determine which reactor, conversion, and radiator technologies would best fulfill future Megawatt Class Nuclear Space Power System Requirements. Specifically, the requirement was 10 megawatts for 5 years of full power operation and 10 years systems life on orbit. A variety of liquid metal and gas cooled reactors, static and dynamic conversion systems, and passive and dynamic radiators were considered. Four concepts were selected for more detailed study. The concepts are: a gas cooled reactor with closed cycle Brayton turbine-alternator conversion with heat pipe and pumped tube-fin heat rejection; a lithium cooled reactor with a free piston Stirling engine-linear alternator and a pumped tube-fin radiator; a lithium cooled reactor with potassium Rankine turbine-alternator and heat pipe radiator; and a lithium cooled incore thermionic static conversion reactor with a heat pipe radiator. The systems recommended for further development to meet a 10 megawatt long life requirement are the lithium cooled reactor with the K-Rankine conversion and heat pipe radiator, and the lithium cooled incore thermionic reactor with heat pipe radiator.

The THERMIE energy farm project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lange, H.J. de; Barbucci, P.; Greil, C.

1998-07-01

At a site close to Pisa (Italy), a plant with a net power-output of approximately 12 MWe will be erected; it features an atmospheric, air-blown, circulating fluidized-bed (CFB) gasifier, integrated with a 10.9 MWe, single-shaft, heavy-duty gas-turbine, suited to burn the low-calorific value fuel-gas produced by the gasifier, and a heat-recovery steam-generator (HRSG), which provides steam to a 5 MWe condensing steam-turbine. The plant's net thermal-efficiency amounts to about 32%. Wet wood is shredded to chips, mixed with the agricultural residues and fed to a dryer. Here, flue gases from the HRSG are used to dry the fuel to themore » desired moisture content. The dried fuel is gasified in a CFB reactor to produce a fuel-gas. This fuel-gas is cooled in two stages during which the gasification air is preheated and steam is produced in a gas-cooler (GC). Then, it is washed in a wet-scrubber and compressed in several intercooled stages before it is delivered to the gas turbine model PGT10 B/1. The gas turbine adopts a newly developed, high-efficiency air-compressor and its special, dual-fuel combustion-chamber is now under development. The combined-cycle is completed with a dual pressure-level HRSG. The steam turbine is fed by the steam produced in the HRSG and the GC located in the gasification island. As a consequence of the changed agricultural market, the fuel will consist not only of short rotation forestry (SRF) but also of forestry and agricultural residues. The wood species include poplar, black locust, willow and chestnut, whereas the agricultural residues comprise olive stones and grape-seed flour. The financial viability of the project relies on the incentive provided by the Italian government with a premium price per kWh of electricity produced from renewables and on the financial contribution from the EU.« less

PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weber, P.; Umminger, K.J.; Schoen, B.

1995-09-01

The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where themore » decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Su'ud, Zaki; Anshari, Rio

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environmentmore » such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

NASA Astrophysics Data System (ADS)

Su'ud, Zaki; Anshari, Rio

2012-06-01

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

THE COOLING REQUIREMENTS AND PROCESS SYSTEMS OF THE SOUTH AFRICAN RESEARCH REACTOR, SAFARI 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Colley, J.R.

1962-12-01

The SAFARI 1 research reactor is cooled and moderated by light water. There are three process systems, a primary water system which cools the reactor core and surroundings, a pool water system, and a secondary water system which removes the heat from the primary and pool systems. The cooling requirements for the reactor core and experimental facilities are outlined, and the cooling and purification functions of the three process systems are described. (auth)

Nuclear reactor melt-retention structure to mitigate direct containment heating

DOEpatents

Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

1991-01-01

A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harmony, S.C.; Steiner, J.L.; Stumpf, H.J.

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

FREE-SURFACE SEPARATION OF STEAM AND WATER FOR APPLICATION IN A MARINE REACTOR AT 1000 PSIG

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steamer, A.G.; Ongman, H.D.

1960-07-13

A series of free-surface steam separation tests were carried out at 1000 psig to obtain data to aid in checking out analytical methods for the effect of ship-s motion on steam separation. Data are presented on the shape and height of the steam-water interface with respect to the indicated water level for two vessel sizes. Further data are presented on the effects of water level and downcomer water velocity on steam carryunder. (auth)

Zirconium alloys with small amounts of iron and copper or nickel show improved corrosion resistance in superheated steam

NASA Technical Reports Server (NTRS)

Greenberg, S.; Youngdahl, C. A.

1967-01-01

Heat treating various compositions of zirconium alloys improve their corrosion resistance to superheated steam at temperatures higher than 500 degrees C. This increases their potential as fuel cladding for superheated-steam nuclear-fueled reactors as well as in autoclaves operating at modest pressures.

Actual operation and regulatory activities on steam generator replacement in Japan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saeki, Hitoshi

1997-02-01

This paper summarizes the operating reactors in Japan, and the status of the steam generators in these plants. It reviews plans for replacement of existing steam generators, and then goes into more detail on the planning and regulatory steps which must be addressed in the process of accomplishing this maintenance. The paper also reviews the typical steps involved in the process of removal and replacement of steam generators.

Reactor water cleanup system

DOEpatents

Gluntz, Douglas M.; Taft, William E.

1994-01-01

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

Research of heat transfer of staggered horizontal bundles of finned tubes at free air convection

NASA Astrophysics Data System (ADS)

Novozhilova, A. V.; Maryna, Z. G.; Samorodov, A. V.; Lvov, E. A.

2017-11-01

The study of free-convective processes is important because of the cooling problem in many machines and systems, where other ways of cooling are impossible or impractical. Natural convective processes are common in the steam turbine air condensers of electric power plants located within the city limits, in dry cooling towers of circulating water systems, in condensers cooled by air and water, in radiators cooling oil of power electric transformers, in emergency cooling systems of nuclear reactors, in solar power, as well as in air-cooling of power semiconductor energy converters. All this makes actual the synthesis of the results of theoretical and experimental research of free convection for heat exchangers with finned tube bundles. The results of the study of free-convection heat transfer for two-, three- and four-row staggered horizontal bundles of industrial bimetallic finned tubes with finning factor of 16.8 and equilateral tubes arrangement are presented. Cross and diagonal steps in the bundles are the same: 58; 61; 64; 70; 76; 86; 100 mm, which corresponds to the relative steps: 1.042; 1.096; 1.152; 1.258; 1.366; 1.545; 1.797. These steps are standardized for air coolers. An equation for calculating the free-convection heat transfer, taking into account the influence of geometrical parameters in the range of Rayleigh number from 30,000 to 350,000 with an average deviation of ± 4.8%, has been obtained. The relationship presented in the article allows designing a wide range of air coolers for various applications, working in the free convection modes.

Concentrating Solar Power Projects - Enerstar | Concentrating Solar Power |

Science.gov Websites

Capacity (Net): 50.0 MW Turbine Manufacturer: Man-Turbo Turbine Description: 3 extractions Output Type : Steam Rankine Power Cycle Pressure: 100.0 bar Cooling Method: Wet cooling Cooling Method Description

Passive cooling system for top entry liquid metal cooled nuclear reactors

DOEpatents

Boardman, Charles E.; Hunsbedt, Anstein; Hui, Marvin M.

1992-01-01

A liquid metal cooled nuclear fission reactor plant having a top entry loop joined satellite assembly with a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during shutdown, or heat produced during a mishap. This satellite type reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary cooling system when rendered inoperative.

Allothermal steam gasification of biomass in cyclic multi-compartment bubbling fluidized-bed gasifier/combustor - new reactor concept.

PubMed

Iliuta, Ion; Leclerc, Arnaud; Larachi, Faïçal

2010-05-01

A new reactor concept of allothermal cyclic multi-compartment fluidized bed steam biomass gasification is proposed and analyzed numerically. The concept combines space and time delocalization to approach an ideal allothermal gasifier. Thermochemical conversion of biomass in periodic time and space sequences of steam biomass gasification and char/biomass combustion is simulated in which the exothermic combustion compartments provide heat into an array of interspersed endothermic steam gasification compartments. This should enhance unit heat integration and thermal efficiency and procure N(2)-free biosyngas with recourse neither to oxygen addition in steam gasification nor contact between flue and syngas. The dynamic, one-dimensional, multi-component, non-isothermal model developed for this concept accounts for detailed solid and gas flow dynamics whereupon gasification/combustion reaction kinetics, thermal effects and freeboard-zone reactions were tied. Simulations suggest that allothermal operation could be achieved with switch periods in the range of a minute supporting practical feasibility for portable small-scale gasification units. Copyright 2009 Elsevier Ltd. All rights reserved.

Analysis on Operating Parameter Design to Steam Methane Reforming in Heat Application RDE

NASA Astrophysics Data System (ADS)

Dibyo, Sukmanto; Sunaryo, Geni Rina; Bakhri, Syaiful; Zuhair; Irianto, Ign. Djoko

2018-02-01

The high temperature reactor has been developed with various power capacities and can produce electricity and heat application. One of heat application is used for hydrogen production. Most hydrogen production occurs by steam reforming that operated at high temperature. This study aims to analyze the feasibility of heat application design of RDE reactor in the steam methane reforming for hydrogen production using the ChemCAD software. The outlet temperature of cogeneration heat exchanger is analyzed to be applied as a feed of steam reformer. Furthermore, the additional heater and calculating amount of fuel usage are described. Results show that at a low mass flow rate of feed, its can produce a temperature up to 480°C. To achieve the temperature of steam methane reforming of 850°C the additional fired heater was required. By the fired heater, an amount of fuel usage is required depending on the Reformer feed temperature produced from the heat exchanger of the cogeneration system.

Glycerol Production and Transformation: A Critical Review with Particular Emphasis on Glycerol Reforming Reaction for Producing Hydrogen in Conventional and Membrane Reactors.

PubMed

Bagnato, Giuseppe; Iulianelli, Adolfo; Sanna, Aimaro; Basile, Angelo

2017-03-23

Glycerol represents an emerging renewable bio-derived feedstock, which could be used as a source for producing hydrogen through steam reforming reaction. In this review, the state-of-the-art about glycerol production processes is reviewed, with particular focus on glycerol reforming reactions and on the main catalysts under development. Furthermore, the use of membrane catalytic reactors instead of conventional reactors for steam reforming is discussed. Finally, the review describes the utilization of the Pd-based membrane reactor technology, pointing out the ability of these alternative fuel processors to simultaneously extract high purity hydrogen and enhance the whole performances of the reaction system in terms of glycerol conversion and hydrogen yield.

NEUTRONIC REACTOR

DOEpatents

Metcalf, H.E.

1957-10-01

A reactor of the type which preferably uses plutonium as the fuel and a liquid moderator, preferably ordinary water, and which produces steam within the reactor core due to the heat of the chain reaction is described. In the reactor shown the fuel elements are essentially in the form of trays and are ventically stacked in spaced relationship. The water moderator is continuously supplied to the trays to maintain a constant level on the upper surfaces of the fuel element as it is continually evaporated by the heat. The steam passes out through the spaces between the fuel elements and is drawn off at the top of the core. The fuel elements are clad in aluminum to prevent deterioration thereof with consequent contamimation of the water.

Glycerol Production and Transformation: A Critical Review with Particular Emphasis on Glycerol Reforming Reaction for Producing Hydrogen in Conventional and Membrane Reactors

PubMed Central

Bagnato, Giuseppe; Iulianelli, Adolfo; Sanna, Aimaro; Basile, Angelo

2017-01-01

Glycerol represents an emerging renewable bio-derived feedstock, which could be used as a source for producing hydrogen through steam reforming reaction. In this review, the state-of-the-art about glycerol production processes is reviewed, with particular focus on glycerol reforming reactions and on the main catalysts under development. Furthermore, the use of membrane catalytic reactors instead of conventional reactors for steam reforming is discussed. Finally, the review describes the utilization of the Pd-based membrane reactor technology, pointing out the ability of these alternative fuel processors to simultaneously extract high purity hydrogen and enhance the whole performances of the reaction system in terms of glycerol conversion and hydrogen yield. PMID:28333121

Laser anemometry measurements of natural circulation flow in a scale model PWR reactor system. [Pressurized Water Reactor

NASA Technical Reports Server (NTRS)

Kadambi, J. R.; Schneider, S. J.; Stewart, W. A.

1986-01-01

The natural circulation of a single phase fluid in a scale model of a pressurized water reactor system during a postulated grade core accident is analyzed. The fluids utilized were water and SF6. The design of the reactor model and the similitude requirements are described. Four LDA tests were conducted: water with 28 kW of heat in the simulated core, with and without the participation of simulated steam generators; water with 28 kW of heat in the simulated core, with the participation of simulated steam generators and with cold upflow of 12 lbm/min from the lower plenum; and SF6 with 0.9 kW of heat in the simulated core and without the participation of the simulated steam generators. For the water tests, the velocity of the water in the center of the core increases with vertical height and continues to increase in the upper plenum. For SF6, it is observed that the velocities are an order of magnitude higher than those of water; however, the velocity patterns are similar.

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 4: Concepts selection, conceptual designs, recommendations

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

A study was conducted by NASA Lewis Research Center for the Triagency SP-100 program office. The objective was to determine which reactor, conversion and radiator technologies would best fulfill future Megawatt Class Nuclear Space Power System Requirements. The requirement was 10 megawatts for 5 years of full power operation and 10 years system life on orbit. A variety of liquid metal and gas cooled reactors, static and dynamic conversion systems, and passive and dynamic radiators were considered. Four concepts were selected for more detailed study: (1) a gas cooled reactor with closed cycle Brayton turbine-alternator conversion with heatpipe and pumped tube fin rejection, (2) a Lithium cooled reactor with a free piston Stirling engine-linear alternator and a pumped tube-fin radiator,(3) a Lithium cooled reactor with a Potassium Rankine turbine-alternator and heat pipe radiator, and (4) a Lithium cooled incore thermionic static conversion reactor with a heat pipe radiator. The systems recommended for further development to meet a 10 megawatt long life requirement are the Lithium cooled reactor with the K-Rankine conversion and heat pipe radiator, and the Lithium cooled incore thermionic reactor with heat pipe radiator.

Reactor water cleanup system

DOEpatents

Gluntz, D.M.; Taft, W.E.

1994-12-20

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR

DOEpatents

Tokarz, R.D.

1981-10-27

This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belles, Randy; Poore, III, Willis P.; Brown, Nicholas R.

2017-03-01

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

The United States Naval Nuclear Propulsion Program - Over 151 Million Miles Safely Steamed on Nuclear Power

DOE Office of Scientific and Technical Information (OSTI.GOV)

None, None

NNSA’s third mission pillar is supporting the U.S. Navy’s ability to protect and defend American interests across the globe. The Naval Reactors Program remains at the forefront of technological developments in naval nuclear propulsion and ensures a commanding edge in warfighting capabilities by advancing new technologies and improvements in naval reactor performance and reliability. In 2015, the Naval Nuclear Propulsion Program pioneered advances in nuclear reactor and warship design – such as increasing reactor lifetimes, improving submarine operational effectiveness, and reducing propulsion plant crewing. The Naval Reactors Program continued its record of operational excellence by providing the technical expertise requiredmore » to resolve emergent issues in the Nation’s nuclear-powered fleet, enabling the Fleet to safely steam more than two million miles. Naval Reactors safely maintains, operates, and oversees the reactors on the Navy’s 82 nuclear-powered warships, constituting more than 45 percent of the Navy’s major combatants.« less

Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

NASA Astrophysics Data System (ADS)

Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-24

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

10 CFR 54.21 - Contents of application-technical information.

Code of Federal Regulations, 2011 CFR

2011-01-01

...), motors, diesel generators, air compressors, snubbers, the control rod drive, ventilation dampers..., the reactor vessel, the reactor coolant system pressure boundary, steam generators, the pressurizer...

10 CFR 54.21 - Contents of application-technical information.

Code of Federal Regulations, 2010 CFR

2010-01-01

...), motors, diesel generators, air compressors, snubbers, the control rod drive, ventilation dampers..., the reactor vessel, the reactor coolant system pressure boundary, steam generators, the pressurizer...

Spectral measurements of direct and scattered gamma radiation at a boiling-water reactor site

NASA Astrophysics Data System (ADS)

Block, R. C.; Preiss, I. L.; Ryan, R. M.; Vargo, G. J.

1990-12-01

Quantitative surveys of direct and scattered gamma radiation emitted from the steam-power conversion systems of a boiling-water reactor and other on-site radiation sources were made using a directionally shielded HPGe gamma spectrometry system. The purpose of this study was to obtain data on the relative contributions and energy distributions of direct and scattered gamma radiation in the site environs. The principal radionuclide of concern in this study is 16N produced by the 16O(n,p) 16N reaction in the reactor coolant. Due to changes in facility operation resulting from the implementation of hydrogen water chemistry (HWC), the amount of 16N transported from the reactor to the main steam system under full power operation is excepted to increase by a factor of 1.2 to 5.0. This increase in the 16N source term in the nuclear steam must be considered in the design of new facilities to be constructed on site as well as the evaluation of existing facilities with repect to ALARA (As Low As Reasonably Achievable) dose limits in unrestricted areas. This study consisted of base-line measurements taken under normal BWR chemistry conditions in October, 1987 and a corresponding set taken under HWC conditions in July, 1988. Ground-level and elevated measurements, corresponding to second-story building height, were obtained. The primary conclusion of this study is that direct radiation from the steam-power conversion system is the predominant source of radiation in the site environs of this reactor and that air scattering (i.e. skyshine) does not appear to be significant.

Steam jacket dynamics in underground coal gasification

NASA Astrophysics Data System (ADS)

Otto, Christopher; Kempka, Thomas

2017-04-01

Underground coal gasification (UCG) has the potential to increase the world-wide hydrocarbon reserves by utilization of deposits not economically mineable by conventional methods. In this context, UCG involves combusting coal in-situ to produce a high-calorific synthesis gas, which can be applied for electricity generation or chemical feedstock production. Apart from high economic potentials, in-situ combustion may cause environmental impacts such as groundwater pollution by by-product leakage. In order to prevent or significantly mitigate these potential environmental concerns, UCG reactors are generally operated below hydrostatic pressure to limit the outflow of UCG process fluids into overburden aquifers. This pressure difference effects groundwater inflow into the reactor and prevents the escape of product gas. In the close reactor vicinity, fluid flow determined by the evolving high reactor temperatures, resulting in the build-up of a steam jacket. Numerical modeling is one of the key components to study coupled processes in in-situ combustion. We employed the thermo-hydraulic numerical simulator MUFITS (BINMIXT module) to address the influence of reactor pressure dynamics as well as hydro-geological coal and caprock parameters on water inflow and steam jacket dynamics. The US field trials Hanna and Hoe Creek (Wyoming) were applied for 3D model validation in terms of water inflow matching, whereby the good agreement between our modeling results and the field data indicates that our model reflects the hydrothermal physics of the process. In summary, our validated model allows a fast prediction of the steam jacket dynamics as well as water in- and outflows, required to avoid aquifer contamination during the entire life cycle of in-situ combustion operations.

Process assessment of small scale low temperature methanol synthesis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hendriyana; Chemical Engineering Department, Faculty of Industrial Technology, InstitutTeknologi Bandung; Susanto, Herri, E-mail: herri@che.itb.ac.id

2015-12-29

Biomass is a renewable energy resource and has the potential to make a significant impact on domestic fuel supplies. Biomass can be converted to fuel like methanol via several step process. The process can be split into following main steps: biomass preparation, gasification, gas cooling and cleaning, gas shift and methanol synthesis. Untill now these configuration still has a problem like high production cost, catalyst deactivation, economy of scale and a huge energy requirements. These problems become the leading inhibition for biomass conversion to methanol, which should be resolved to move towards the economical. To address these issues, we developedmore » various process and new configurations for methanol synthesis via methyl formate. This configuration combining two reactors: the one reactor for the carbonylation of methanol and CO to form methyl formate, and the second for the hydrogenolysis of methyl formate and H{sub 2} to form two molecule of methanol. Four plant process configurations were compared with the biomass basis is 300 ton/day. The first configuration (A) is equipped with a steam reforming process for converting methane to CO and H{sub 2} for increasing H{sub 2}/CO ratio. CO{sub 2} removal is necessary to avoid poisoning the catalyst. COSORB process used for the purpose of increasing the partial pressure of CO in the feed gas. The steam reforming process in B configuration is not used with the aim of reducing the number of process equipment, so expect lower investment costs. For C configuration, the steam reforming process and COSORB are not used with the aim of reducing the number of process equipment, so expect lower investment costs. D configuration is almost similar to the configuration A. This configuration difference is in the synthesis of methanol which was held in a single reactor. Carbonylation and hydrogenolysis reactions carried out in the same reactor one. These processes were analyzed in term of technical process, material and energy balance and economic analysis. The presented study is an attempt to compile most of these efforts in order to guide future work to get cheaper low cost investment. From our study the interesting configuration to the next development is D configuration with methanol yield 112 ton/day and capital cost with 526.4 10{sup 6} IDR. The configuration of D with non-discounted and discounted rate had the break-even point approximately six and eight years.« less

Hydrogen production by reforming of liquid hydrocarbons in a membrane reactor for portable power generation-Experimental studies

NASA Astrophysics Data System (ADS)

Damle, Ashok S.

One of the most promising technologies for lightweight, compact, portable power generation is proton exchange membrane (PEM) fuel cells. PEM fuel cells, however, require a source of pure hydrogen. Steam reforming of hydrocarbons in an integrated membrane reactor has potential to provide pure hydrogen in a compact system. Continuous separation of product hydrogen from the reforming gas mixture is expected to increase the yield of hydrogen significantly as predicted by model simulations. In the laboratory-scale experimental studies reported here steam reforming of liquid hydrocarbon fuels, butane, methanol and Clearlite ® was conducted to produce pure hydrogen in a single step membrane reformer using commercially available Pd-Ag foil membranes and reforming/WGS catalysts. All of the experimental results demonstrated increase in hydrocarbon conversion due to hydrogen separation when compared with the hydrocarbon conversion without any hydrogen separation. Increase in hydrogen recovery was also shown to result in corresponding increase in hydrocarbon conversion in these studies demonstrating the basic concept. The experiments also provided insight into the effect of individual variables such as pressure, temperature, gas space velocity, and steam to carbon ratio. Steam reforming of butane was found to be limited by reaction kinetics for the experimental conditions used: catalysts used, average gas space velocity, and the reactor characteristics of surface area to volume ratio. Steam reforming of methanol in the presence of only WGS catalyst on the other hand indicated that the membrane reactor performance was limited by membrane permeation, especially at lower temperatures and lower feed pressures due to slower reconstitution of CO and H 2 into methane thus maintaining high hydrogen partial pressures in the reacting gas mixture. The limited amount of data collected with steam reforming of Clearlite ® indicated very good match between theoretical predictions and experimental results indicating that the underlying assumption of the simple model of conversion of hydrocarbons to CO and H 2 followed by equilibrium reconstitution to methane appears to be reasonable one.

1170 MW/sub t/ HTGR steamer cogeneration plant: design and cost study

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

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 numbermore » of basic tasks such as the development of plant concept, capital cost estimate, project schedule and annual operation and maintenance (O and M) cost.« less

The first PANDA tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dreier, J.; Huggenberger, M.; Aubert, C.

1996-08-01

The PANDA test facility at PSI in Switzerland is used to study the long-term Simplified Boiling Water Reactor (SBWR) Passive Containment Cooling System (PCCS) performance. The PANDA tests demonstrate performance on a larger scale than previous tests and examine the effects of any non-uniform spatial distributions of steam and non-condensables in the system. The PANDA facility has a 1:1 vertical scale, and 1:25 ``system`` scale (volume, power, etc.). Steady-state PCCS condenser performance tests and extensive facility characterization tests have been completed. Transient system behavior tests were conducted late in 1995; results from the first three transient tests (M3 series) aremore » reviewed. The first PANDA tests showed that the overall global behavior of the SBWR containment was globally repeatable and very favorable; the system exhibited great ``robustness.``« less

Process for purifying geothermal steam

DOEpatents

Li, Charles T.

1980-01-01

Steam containing hydrogen sulfide is purified and sulfur recovered by passing the steam through a reactor packed with activated carbon in the presence of a stoichiometric amount of oxygen which oxidizes the hydrogen sulfide to elemental sulfur which is adsorbed on the bed. The carbon can be recycled after the sulfur has been recovered by vacuum distillation, inert gas entrainment or solvent extraction. The process is suitable for the purification of steam from geothermal sources which may also contain other noncondensable gases.

Process for purifying geothermal steam

DOEpatents

Li, C.T.

Steam containing hydrogen sulfide is purified and sulfur recovered by passing the steam through a reactor packed with activated carbon in the presence of a stoichiometric amount of oxygen which oxidizes the hydrogen sulfide to elemental sulfur which is adsorbed on the bed. The carbon can be recycled after the sulfur has been recovered by vacuum distillation, inert gas entrainment or solvent extraction. The process is suitable for the purification of steam from geothermal sources which may also contain other noncondensable gases.

Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 2: Application to EBR-II Primary Sodium System and Related Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steven R. Sherman; Collin J. Knight

Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/ormore » to comply with decontamination and decomissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidifed carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, USA. This report is Part 2 of a two-part report. This second report provides a supplement to the first report and describes the application of the humdidified carbon dioxide technique ("carbonation") to the EBR-II primary tank, primary cover gas systems, and the intermediate heat exchanger. Future treatment plans are also provided.« less

Use of joint-growth directions and rock textures to infer thermal regimes during solidification of basaltic lava flows

NASA Astrophysics Data System (ADS)

Degraff, James M.; Long, Philip E.; Aydin, Atilla

1989-09-01

Thermal contraction joints form in the upper and lower solidifying crusts of basaltic lava flows and grow toward the interior as the crusts thicken. Lava flows are thus divided by vertical joints that, by changes in joint spacing and form, define horizontal intraflow layers known as tiers. Entablatures are tiers with joint spacings less than about 40 cm, whereas colonnades have larger joint spacings. We use structural and petrographic methods to infer heat-transfer processes and to constrain environmental conditions that produce these contrasting tiers. Joint-surface morphology indicates overall joint-growth direction and thus identifies the level in a flow where the upper and lower crusts met. Rock texture provides information on relative cooling rates in the tiers of a flow. Lava flows without entablature have textures that develop by relatively slow cooling, and two joint sets that usually meet near their middles, which indicate mostly conductive cooling. Entablature-bearing flows have two main joint sets that meet well below their middles, and textures that indicate fast cooling of entablatures and slow cooling of colonnades. Entablatures always occur in the upper joint sets and sometimes alternate several times with colonnades. Solidification times of entablature-bearing flows, constrained by lower joint-set thicknesses, are much less than those predicted by a purely conductive cooling model. These results are best explained by a cooling model based on conductive heat transfer near a flow base and water-steam convection in the upper part of an entablature-bearing flow. Calculated solidification rates in the upper parts of such flows exceed that of the upper crust of Kilauea Iki lava lake, where water-steam convection is documented. Use of the solidification rates in an available model of water-steam convection yields permeability values that agree with measured values for fractured crystalline rock. We conclude, therefore, that an entablature forms when part of a flow cools very rapidly by water-steam convection. Flooding of the flow top by surface drainage most likely induces the convection. Colonnades form under conditions of slower cooling by conductive heat transfer in the absence of water.

Analysis of boron dilution in a four-loop PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sun, J.G.; Sha, W.T.

1995-12-31

Thermal mixing and boron dilution in a pressurized water reactor were analyzed with COMMIX codes. The reactor system was the four loop Zion reactor. Two boron dilution scenarios were analyzed. In the first scenario, the plant is in cold shutdown and the reactor coolant system has just been filled after maintenance on the steam generators. To flush the air out of the steam generator tubes, a reactor coolant pump (RCP) is started, with the water in the pump suction line devoid of boron and at the same temperature as the coolant in the system. In the second scenario, the plantmore » is at hot standby and the reactor coolant system has been heated up to operating temperature after a long outage. It is assumed that an RCP is started, with the pump suction line filled with cold unborated water, forcing a slug of diluted coolant down the downcomer and subsequently through the reactor core. The subsequent transient thermal mixing and boron dilution that would occur in the reactor system is simulated for these two scenarios. The reactivity insertion rate and the total reactivity are evaluated.« less

Implementing a Nuclear Power Plant Model for Evaluating Load-Following Capability on a Small Grid

NASA Astrophysics Data System (ADS)

Arda, Samet Egemen

A pressurized water reactor (PWR) nuclear power plant (NPP) model is introduced into Positive Sequence Load Flow (PSLF) software by General Electric in order to evaluate the load-following capability of NPPs. The nuclear steam supply system (NSSS) consists of a reactor core, hot and cold legs, plenums, and a U-tube steam generator. The physical systems listed above are represented by mathematical models utilizing a state variable lumped parameter approach. A steady-state control program for the reactor, and simple turbine and governor models are also developed. Adequacy of the isolated reactor core, the isolated steam generator, and the complete PWR models are tested in Matlab/Simulink and dynamic responses are compared with the test results obtained from the H. B. Robinson NPP. Test results illustrate that the developed models represents the dynamic features of real-physical systems and are capable of predicting responses due to small perturbations of external reactivity and steam valve opening. Subsequently, the NSSS representation is incorporated into PSLF and coupled with built-in excitation system and generator models. Different simulation cases are run when sudden loss of generation occurs in a small power system which includes hydroelectric and natural gas power plants besides the developed PWR NPP. The conclusion is that the NPP can respond to a disturbance in the power system without exceeding any design and safety limits if appropriate operational conditions, such as achieving the NPP turbine control by adjusting the speed of the steam valve, are met. In other words, the NPP can participate in the control of system frequency and improve the overall power system performance.

Apparatus and method for solar coal gasification

DOEpatents

Gregg, David W.

1980-01-01

Apparatus for using focused solar radiation to gasify coal and other carbonaceous materials. Incident solar radiation is focused from an array of heliostats onto a tower-mounted secondary mirror which redirects the focused solar radiation down through a window onto the surface of a vertically-moving bed of coal, or a fluidized bed of coal, contained within a gasification reactor. The reactor is designed to minimize contact between the window and solids in the reactor. Steam introduced into the gasification reactor reacts with the heated coal to produce gas consisting mainly of carbon monoxide and hydrogen, commonly called "synthesis gas", which can be converted to methane, methanol, gasoline, and other useful products. One of the novel features of the invention is the generation of process steam at the rear surface of the secondary mirror.

Apparatus for solar coal gasification

DOEpatents

Gregg, D.W.

Apparatus for using focused solar radiation to gasify coal and other carbonaceous materials is described. Incident solar radiation is focused from an array of heliostats onto a tower-mounted secondary mirror which redirects the focused solar radiation down through a window onto the surface of a vertically-moving bed of coal, or a fluidized bed of coal, contained within a gasification reactor. The reactor is designed to minimize contact between the window and solids in the reactor. Steam introduced into the gasification reactor reacts with the heated coal to produce gas consisting mainly of carbon monoxide and hydrogen, commonly called synthesis gas, which can be converted to methane, methanol, gasoline, and other useful products. One of the novel features of the invention is the generation of process steam at the rear surface of the secondary mirror.

Steam reforming of heptane in a fluidized bed membrane reactor

NASA Astrophysics Data System (ADS)

Rakib, Mohammad A.; Grace, John R.; Lim, C. Jim; Elnashaie, Said S. E. H.

n-Heptane served as a model compound to study steam reforming of naphtha as an alternative feedstock to natural gas for production of pure hydrogen in a fluidized bed membrane reactor. Selective removal of hydrogen using Pd 77Ag 23 membrane panels shifted the equilibrium-limited reactions to greater conversion of the hydrocarbons and lower yields of methane, an intermediate product. Experiments were conducted with no membranes, with one membrane panel, and with six panels along the height of the reactor to understand the performance improvement due to hydrogen removal in a reactor where catalyst particles were fluidized. Results indicate that a fluidized bed membrane reactor (FBMR) can provide a compact reformer for pure hydrogen production from a liquid hydrocarbon feedstock at moderate temperatures (475-550 °C). Under the experimental conditions investigated, the maximum achieved yield of pure hydrogen was 14.7 moles of pure hydrogen per mole of heptane fed.

Interim status report on lead-cooled fast reactor (LFR) research and development.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.

2008-03-31

This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigationmore » of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup 15} (n/cm{sup 2}-s) and the initially 563 MWt PHENIX reactor attained 2.0 x 10{sup 15} (n/cm{sup 2}-s) before one of three intermediate cooling loops was shut down due to concerns about potential steam generator tube failures. The calculations do not assume a test assembly location for advanced fuels and materials irradiation in place of a fuel assembly (e.g., at the center of the core); the calculations have not examined whether it would be feasible to replace the central assembly by a test assembly location. However, having only fifteen driver assemblies implies a significant effect due to perturbations introduced by the test assembly. The peak neutron fast flux is low compared with the fast fluxes previously achieved in FFTF and PHENIX. Furthermore, the peak neutron fluence is only about half of the limiting value (4 x 10{sup 23} n/cm{sup 2}) typically used for ferritic steels. The results thus suggest that a larger power level (e.g., 400 MWt) and a larger core would be better for a TPP based upon the ELSY fuel assembly design and which can also perform irradiation testing of advanced fuels and materials. In particular, a core having a higher power level and larger dimensions would achieve a suitable average discharge burnup, peak fast flux, peak fluence, and would support the inclusion of one or more test assembly locations. Participation in the Generation IV International Forum Provisional System Steering Committee for the LFR is being maintained throughout FY 2008. Results from the analysis of samples previously exposed to flowing lead-bismuth eutectic (LBE) in the DELTA loop are summarized and a model for the oxidation/corrosion kinetics of steels in heavy liquid metal coolants was applied to systematically compare the calculated long-term (i.e., following several years of growth) oxide layer thicknesses of several steels.« less

High yields of hydrogen production from methanol steam reforming with a cross-U type reactor

PubMed Central

Zhang, Shubin; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

2017-01-01

This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance. PMID:29121067

High yields of hydrogen production from methanol steam reforming with a cross-U type reactor.

PubMed

Zhang, Shubin; Zhang, Yufeng; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

2017-01-01

This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance.

Energy Conversion Alternatives Study (ECAS), Westinghouse phase 1. Volume 9: Closed-cycle MHD. [energy conversion efficiency of electric power plants using magnetohydrodynamics

NASA Technical Reports Server (NTRS)

Tsu, T. C.

1976-01-01

A closed-cycle MHD system for an electric power plant was studied. It consists of 3 interlocking loops, an external heating loop, a closed-cycle cesium seeded argon nonequilibrium ionization MHD loop, and a steam bottomer. A MHD duct maximum temperature of 2366 K (3800 F), a pressure of 0.939 MPa (9.27 atm) and a Mach number of 0.9 are found to give a topping cycle efficiency of 59.3%; however when combined with an integrated gasifier and optimistic steam bottomer the coal to bus bar efficiency drops to 45.5%. A 1978 K (3100 F) cycle has an efficiency of 55.1% and a power plant efficiency of 42.2%. The high cost of the external heating loop components results in a cost of electricity of 21.41 mills/MJ (77.07 mills/kWh) for the high temperature system and 19.0 mills/MJ (68.5 mills/kWh) for the lower temperature system. It is, therefore, thought that this cycle may be more applicable to internally heated systems such as some futuristic high temperature gas cooled reactor.

Phase separation of metal-added corium and its effect on a steam explosion

NASA Astrophysics Data System (ADS)

Min, B. T.; Kim, J. H.; Hong, S. W.; Hong, S. H.; Park, I. K.; Song, J. H.; Kim, H. D.

2008-07-01

To simulate a relocation of molten core material and its interaction phenomenon with water during a severe accident in a nuclear reactor, a typical corium of UO 2/ZrO 2/Zr/Stainless steel mixed at a 62 wt%, 15 wt%, 12 wt% and 11 wt%, respectively, was melted and then cooled down to become a solidified ingot. It was shown that the molten corium was separated into two layers, of which the upper layer was oxide mixtures and the lower layer was metal alloys. The upper layer was UO 2 and ZrO 2 and the lower layer mostly consisted of metal mixtures such as uranium, zirconium and stainless steel. Iron content varied with the positions and about a half of it existed as an alloy such as Fe 2U. Uranium metal was produced by reduction of UO 2 by zirconium metal. The average densities of the upper oxide layer and the lower metal layer were 8.802 and 9.411 g/cm 3, respectively. In another test, metal-added molten corium was poured into water and it showed that a steam explosion could occur by applying an external trigger.

CHAIN REACTING SYSTEM

DOEpatents

Fermi, E.; Leverett, M.C.

1958-06-01

A nuclear reactor of the gas-cooled, graphitemoderated type is described. In this design, graphite blocks are arranged in a substantially cylindrical lattice having vertically orienied coolant channels in which uranium fuel elements having through passages are disposed. The active lattice is contained within a hollow body. such as a steel shell, which, in turn, is surrounded by water and concrete shields. Helium is used as the primary coolant and is circulated under pressure through the coolant channels and fuel elements. The helium is then conveyed to heat exchangers, where its heat is used to produce steam for driving a prime mover, thence to filtering means where radioactive impurities are removed. From the filtering means the helium passes to a compressor and an after cooler and is ultimately returned to the reactor for recirculation. Control and safety rods are provided to stabilize or stop the reaction. A space is provided between the graphite lattice and the internal walls of the shell to allow for thermal expansion of the lattice during operation. This space is filled with a resilient packing, such as asbestos, to prevent the passage of helium.

Application of waterproof breathable fabric in thermal protective clothing exposed to hot water and steam

NASA Astrophysics Data System (ADS)

Su, Y.; Li, R.; Song, G.; Li, J.

2017-10-01

A hot water and steam tester was used to examine thermal protective performance of waterproof and breathable fabric against hot water and steam hazards. Time to cause skin burn and thermal energy absorbed by skin during exposure and cooling phases was employed to characterize the effect of configuration, placing order and properties of waterproof and breathable fabric on the thermal protective performance. The difference of thermal protective performance due to hot water and steam hazards was discussed. The result showed that the configuration of waterproof and breathable fabric presented a significant effect on the thermal protective performance of single- and double-layer fabric system, while the difference between different configurations in steam hazard was greater than that in hot water hazard. The waterproof and breathable fabric as outer layer provided better protection than that as inner layer. Increasing thickness and moisture regain improved the thermal protective performance of fabric system. Additionally, the thermal energy absorbed by skin during the cooling phase was affected by configuration, thickness and moisture regain of fabric. The findings will provide technical data to improve performance of thermal protective clothing in hot water and steam hazards.

Circulating moving bed system for CO.sub.2 separation, and method of same

DOEpatents

Elliott, Jeannine Elizabeth; Copeland, Robert James

2016-12-27

A circulating moving bed and process for separating a carbon dioxide from a gas stream is disclosed. The circulating moving bed can include an adsorption reactor and a desorption reactor, and a sorbent that moves through the two reactors. The sorbent can enter the adsorptive reactor and one end and move to an exit point distal to its entry point, while a CO.sub.2 feed stream can enter near the distal point and move countercurrently through the sorbent to exit at a position near the entry point of the sorbent. The sorbent can adsorb the CO.sub.2 by concentration swing adsorption and adsorptive displacement. The sorbent can then transfer to a regeneration reactor and can move countercurrently against a flow of steam through the regeneration reactor. The sorbent can be regenerated and the carbon dioxide recaptured by desorbing the carbon dioxide from the sorbent using concentration swing desorption and desorptive displacement with steam.

Conceptual Design of Low-Temperature Hydrogen Production and High-Efficiency Nuclear Reactor Technology

NASA Astrophysics Data System (ADS)

Fukushima, Kimichika; Ogawa, Takashi

Hydrogen, a potential alternative energy source, is produced commercially by methane (or LPG) steam reforming, a process that requires high temperatures, which are produced by burning fossil fuels. However, as this process generates large amounts of CO2, replacement of the combustion heat source with a nuclear heat source for 773-1173K processes has been proposed in order to eliminate these CO2 emissions. In this paper, a novel method of nuclear hydrogen production by reforming dimethyl ether (DME) with steam at about 573K is proposed. From a thermodynamic equilibrium analysis of DME steam reforming, the authors identified conditions that provide high hydrogen production fraction at low pressure and temperatures of about 523-573K. By setting this low-temperature hydrogen production process upstream from a turbine and nuclear reactor at about 573K, the total energy utilization efficiency according to equilibrium mass and heat balance analysis is about 50%, and it is 75%for a fast breeder reactor (FBR), where turbine is upstream of the reformer.

Reforming of natural gas—hydrogen generation for small scale stationary fuel cell systems

NASA Astrophysics Data System (ADS)

Heinzel, A.; Vogel, B.; Hübner, P.

The reforming of natural gas to produce hydrogen for fuel cells is described, including the basic concepts (steam reforming or autothermal reforming) and the mechanisms of the chemical reactions. Experimental work has been done with a compact steam reformer, and a prototype of an experimental reactor for autothermal reforming was tested, both containing a Pt-catalyst on metallic substrate. Experimental results on the steam reforming system and a comparison of the steam reforming process with the autothermal process are given.

System Modeling for Ammonia Synthesis Energy Recovery System

NASA Astrophysics Data System (ADS)

Bran Anleu, Gabriela; Kavehpour, Pirouz; Lavine, Adrienne; Ammonia thermochemical Energy Storage Team

2015-11-01

An ammonia thermochemical energy storage system is an alternative solution to the state-of-the-art molten salt TES system for concentrating solar power. Some of the advantages of this emerging technology include its high energy density, no heat losses during the storage duration, and the possibility of long storage periods. Solar energy powers an endothermic reaction to disassociate ammonia into hydrogen and nitrogen, which can be stored for future use. The reverse reaction is carried out in the energy recovery process; a hydrogen-nitrogen mixture flowing through a catalyst bed undergoes the exothermic ammonia synthesis reaction. The goal is to use the ammonia synthesis reaction to heat supercritical steam to temperatures on the order of 650°C as required for a supercritical steam Rankine cycle. The steam will flow through channels in a combined reactor-heat exchanger. A numerical model has been developed to determine the optimal design to heat supercritical steam while maintaining a stable exothermic reaction. The model consists of a transient one dimensional concentric tube counter-flow reactor-heat exchanger. The numerical model determines the inlet mixture conditions needed to achieve various steam outlet conditions.

Dry Air Cooler Modeling for Supercritical Carbon Dioxide Brayton Cycle Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moisseytsev, A.; Sienicki, J. J.; Lv, Q.

Modeling for commercially available and cost effective dry air coolers such as those manufactured by Harsco Industries has been implemented in the Argonne National Laboratory Plant Dynamics Code for system level dynamic analysis of supercritical carbon dioxide (sCO 2) Brayton cycles. The modeling can now be utilized to optimize and simulate sCO 2 Brayton cycles with dry air cooling whereby heat is rejected directly to the atmospheric heat sink without the need for cooling towers that require makeup water for evaporative losses. It has sometimes been stated that a benefit of the sCO 2 Brayton cycle is that it enablesmore » dry air cooling implying that the Rankine steam cycle does not. A preliminary and simple examination of a Rankine superheated steam cycle and an air-cooled condenser indicates that dry air cooling can be utilized with both cycles provided that the cycle conditions are selected appropriately« less

Mathematical Modeling – The Impact of Cooling Water Temperature Upsurge on Combined Cycle Power Plant Performance and Operation

NASA Astrophysics Data System (ADS)

Indra Siswantara, Ahmad; Pujowidodo, Hariyotejo; Darius, Asyari; Ramdlan Gunadi, Gun Gun

2018-03-01

This paper presents the mathematical modeling analysis on cooling system in a combined cycle power plant. The objective of this study is to get the impact of cooling water upsurge on plant performance and operation, using Engineering Equation Solver (EES™) tools. Power plant installed with total power capacity of block#1 is 505.95 MWe and block#2 is 720.8 MWe, where sea water consumed as cooling media at two unit condensers. Basic principle of analysis is heat balance calculation from steam turbine and condenser, concern to vacuum condition and heat rate values. Based on the result shown graphically, there were impact the upsurge of cooling water to increase plant heat rate and vacuum pressure in condenser so ensued decreasing plant efficiency and causing possibility steam turbine trip as back pressure raised from condenser.

SINGLE PHASE ANALYTICAL MODELS FOR TERRY TURBINE NOZZLE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

All BWR RCIC (Reactor Core Isolation Cooling) systems and PWR AFW (Auxiliary Feed Water) systems use Terry turbine, which is composed of the wheel with turbine buckets and several groups of fixed nozzles and reversing chambers inside the turbine casing. The inlet steam is accelerated through the turbine nozzle and impacts on the wheel buckets, generating work to drive the RCIC pump. As part of the efforts to understand the unexpected “self-regulating” mode of the RCIC systems in Fukushima accidents and extend BWR RCIC and PWR AFW operational range and flexibility, mechanistic models for the Terry turbine, based on Sandiamore » National Laboratories’ original work, has been developed and implemented in the RELAP-7 code to simulate the RCIC system. RELAP-7 is a new reactor system code currently under development with the funding support from U.S. Department of Energy. The RELAP-7 code is a fully implicit code and the preconditioned Jacobian-free Newton-Krylov (JFNK) method is used to solve the discretized nonlinear system. This paper presents a set of analytical models for simulating the flow through the Terry turbine nozzles when inlet fluid is pure steam. The implementation of the models into RELAP-7 will be briefly discussed. In the Sandia model, the turbine bucket inlet velocity is provided according to a reduced-order model, which was obtained from a large number of CFD simulations. In this work, we propose an alternative method, using an under-expanded jet model to obtain the velocity and thermodynamic conditions for the turbine bucket inlet. The models include both adiabatic expansion process inside the nozzle and free expansion process out of the nozzle to reach the ambient pressure. The combined models are able to predict the steam mass flow rate and supersonic velocity to the Terry turbine bucket entrance, which are the necessary input conditions for the Terry Turbine rotor model. The nozzle analytical models were validated with experimental data and benchmarked with CFD simulations. The analytical models generally agree well with the experimental data and CFD simulations. The analytical models are suitable for implementation into a reactor system analysis code or severe accident code as part of mechanistic and dynamical models to understand the RCIC behaviors. The cases with two-phase flow at the turbine inlet will be pursued in future work.« less

BOILING WATER REACTOR WITH FEED WATER INJECTION NOZZLES

DOEpatents

Treshow, M.

1963-04-30

This patent covers the use of injection nozzles for pumping water into the lower ends of reactor fuel tubes in which water is converted directly to steam. Pumping water through fuel tubes of this type of boiling water reactor increases its power. The injection nozzles decrease the size of pump needed, because the pump handles only the water going through the nozzles, additional water being sucked into the tubes by the nozzles independently of the pump from the exterior body of water in which the fuel tubes are immersed. The resulting movement of exterior water along the tubes holds down steam formation, and thus maintains the moderator effectiveness, of the exterior body of water. (AEC)

Method and apparatus for producing synthesis gas

DOEpatents

Hemmings, John William; Bonnell, Leo; Robinson, Earl T.

2010-03-03

A method and apparatus for reacting a hydrocarbon containing feed stream by steam methane reforming reactions to form a synthesis gas. The hydrocarbon containing feed is reacted within a reactor having stages in which the final stage from which a synthesis gas is discharged incorporates expensive high temperature materials such as oxide dispersed strengthened metals while upstream stages operate at a lower temperature allowing the use of more conventional high temperature alloys. Each of the reactor stages incorporate reactor elements having one or more separation zones to separate oxygen from an oxygen containing feed to support combustion of a fuel within adjacent combustion zones, thereby to generate heat to support the endothermic steam methane reforming reactions.

Fischer-Tropsch Wastewater Utilization

DOEpatents

Shah, Lalit S.

2003-03-18

The present invention is generally directed to handling the wastewater, or condensate, from a hydrocarbon synthesis reactor. More particularly, the present invention provides a process wherein the wastewater of a hydrocarbon synthesis reactor, such as a Fischer-Tropsch reactor, is sent to a gasifier and subsequently reacted with steam and oxygen at high temperatures and pressures so as to produce synthesis gas. The wastewater may also be recycled back to a slurry preparation stage, where solid combustible organic materials are pulverized and mixed with process water and the wastewater to form a slurry, after which the slurry fed to a gasifier where it is reacted with steam and oxygen at high temperatures and pressures so as to produce synthesis gas.

The effectiveness of using the combined-cycle technology in a nuclear power plant unit equipped with an SVBR-100 reactor

NASA Astrophysics Data System (ADS)

Kasilov, V. F.; Dudolin, A. A.; Gospodchenkov, I. V.

2015-05-01

The design of a modular SVBR-100 reactor with a lead-bismuth alloy liquid-metal coolant is described. The basic thermal circuit of a power unit built around the SVBR-100 reactor is presented together with the results of its calculation. The gross electrical efficiency of the turbine unit driven by saturated steam at a pressure of 6.7 MPa is estimated at η{el/gr} = 35.5%. Ways for improving the efficiency of this power unit and increasing its power output by applying gas-turbine and combined-cycle technologies are considered. With implementing a combined-cycle power-generating system comprising two GE-6101FA gas-turbine units with a total capacity of 140 MW, it becomes possible to obtain the efficiency of the combined-cycle plant equipped with the SVBR-100 reactor η{el/gr} = 45.39% and its electrical power output equal to 328 MW. The heat-recovery boiler used as part of this power installation generates superheated steam with a temperature of 560°C, due to which there is no need to use a moisture separator/steam reheater in the turbine unit thermal circuit.

Safety margins in zircaloy oxidation and embrittlement criteria for emergency core cooling system acceptance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williford, R.E.

1986-09-01

Current emergency core cooling system acceptance criteria for light water reactors specify that, under loss-of-coolant accident (LOCA) conditions, the Baker-Just (BJ) correlation must be used to calculate Zircaloy-steam oxidation, calculated peak cladding temperatures (PCT) must not exceed 1204/sup 0/C, and calculated oxidation must not exceed 17% equivalent cladding reacted (ECR). An appropriately defined minimum margin of safety was estimated for each of these criteria. The currently required BJ oxidation correlation provides margins only over the 1100 to 1500/sup 0/C temperature range at the 95% confidence level. The PCT margins for thermal shock and handling failures are adequate at oxidation temperaturesmore » above 1204/sup 0/C for up to 210 and 160 s, respectively, at the 95% confidence level. The ECR thermal shock and handling margins at the 50 and 95% confidence levels, respectively, range between 2 and 7% ECR for the BJ correlation, but vanish at temperatures above 1100 to 1160/sup 0/C for the best-estimate Cathcart-Pawel correlation. However, use of the Cathcart Pawel correlation for ''design basis'' LOCA calculations can be justified at the 85 to 88% confidence level if cooling rate effects can be neglected.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bolstad, J.W.; Haarman, R.A.

The results of two transients involving the loss of a steam generator in a single-pass, steam generator, pressurized water reactor have been analyzed using a state-of-the-art, thermal-hydraulic computer code. Computed results include the formation of a steam bubble in the core while the pressurizer is solid. Calculations show that continued injection of high pressure water would have stopped the scenario. These are similar to the happenings at Three Mile Island.

Improved vortex reactor system

DOEpatents

Diebold, James P.; Scahill, John W.

1995-01-01

An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

The advantages and disadvantages of using the TREAT reactor for nuclear laser experiments

NASA Astrophysics Data System (ADS)

Dickson, P. W.; Snyder, A. M.; Imel, G. R.; McConnell, R. J.

The Transient Reactor Test Facility (TREAT) is a large air-cooled test facility located at the Idaho National Engineering Laboratory. Two of the major design features of TREAT, its large size and its being an air-cooled reactor, provide clues to both its advantages and disadvantages for supporting nuclear laser experiments. Its large size, which is dictated by the dilute uranium/graphite fuel, permits accommodation of geometrically large experiments. However, TREAT's large size also results in relatively long transients so that the energy deposited in an experiment is large relative to the peak power available from the reactor. TREAT's air-cooling mode of operation allows its configuration to be changed fairly readily. Due to air cooling, the reactor cools down slowly, permitting only one full power transient a day, which can be a disadvantage in some experimental programs. The reactor is capable of both steady-state or transient operation.

Pressurized thermal shock: TEMPEST computer code simulation of thermal mixing in the cold leg and downcomer of a pressurized water reactor. [Creare 61 and 64

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eyler, L.L.; Trent, D.S.

The TEMPEST computer program was used to simulate fluid and thermal mixing in the cold leg and downcomer of a pressurized water reactor under emergency core cooling high-pressure injection (HPI), which is of concern to the pressurized thermal shock (PTS) problem. Application of the code was made in performing an analysis simulation of a full-scale Westinghouse three-loop plant design cold leg and downcomer. Verification/assessment of the code was performed and analysis procedures developed using data from Creare 1/5-scale experimental tests. Results of three simulations are presented. The first is a no-loop-flow case with high-velocity, low-negative-buoyancy HPI in a 1/5-scale modelmore » of a cold leg and downcomer. The second is a no-loop-flow case with low-velocity, high-negative density (modeled with salt water) injection in a 1/5-scale model. Comparison of TEMPEST code predictions with experimental data for these two cases show good agreement. The third simulation is a three-dimensional model of one loop of a full size Westinghouse three-loop plant design. Included in this latter simulation are loop components extending from the steam generator to the reactor vessel and a one-third sector of the vessel downcomer and lower plenum. No data were available for this case. For the Westinghouse plant simulation, thermally coupled conduction heat transfer in structural materials is included. The cold leg pipe and fluid mixing volumes of the primary pump, the stillwell, and the riser to the steam generator are included in the model. In the reactor vessel, the thermal shield, pressure vessel cladding, and pressure vessel wall are thermally coupled to the fluid and thermal mixing in the downcomer. The inlet plenum mixing volume is included in the model. A 10-min (real time) transient beginning at the initiation of HPI is computed to determine temperatures at the beltline of the pressure vessel wall.« less

Pyrometer mount for a closed-circuit thermal medium cooled gas turbine

DOEpatents

Jones, Raymond Joseph; Kirkpatrick, Francis Lawrence; Burns, James Lee; Fulton, John Robert

2002-01-01

A steam-cooled second-stage nozzle segment has an outer band and an outer cover defining a plenum therebetween for receiving cooling steam for flow through the nozzles to the inner band and cover therefor and return flow through the nozzles. To measure the temperature of the buckets of the stage forwardly of the nozzle stage, a pyrometer boss is electron beam-welded in an opening through the outer band and TIG-welded to the outer cover plate. By machining a hole through the boss and seating a linearly extending tube in the boss, a line of sight between a pyrometer mounted on the turbine frame and the buckets is provided whereby the temperature of the buckets can be ascertained. The welding of the boss to the outer band and outer cover enables steam flow through the plenum without leakage, while providing a line of sight through the outer cover and outer band to measure bucket temperature.

Investigation of a para-ortho hydrogen reactor for application to spacecraft sensor cooling

NASA Technical Reports Server (NTRS)

Nast, T. C.

1983-01-01

The utilization of solid hydrogen in space for sensor and instrument cooling is a very efficient technique for long term cooling or for cooling at high heat rates. The solid hydrogen can provide temperatures as low as 7 to 8 K to instruments. Vapor cooling is utilized to reduce parasitic heat inputs to the 7 to 8 K stage and is effective in providing intermediate cooling for instrument components operating at higher temperatures. The use of solid hydrogen in place of helium may lead to weight reductions as large as a factor of ten and an attendent reduction in system volume. The results of an investigation of a catalytic reactor for use with a solid hydrogen cooling system is presented. Trade studies were performed on several configurations of reactor to meet the requirements of high reactor efficiency with low pressure drop. Results for the selected reactor design are presented for both liquid hydrogen systems operating at near atmospheric pressure and the solid hydrogen cooler operating as low as 1 torr.

Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

DOEpatents

Hunsbedt, A.; Boardman, C.E.

1995-04-11

A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor is disclosed. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo`s structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated. 5 figures.

Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1995-01-01

A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo's structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated.

METHOD AND APPARATUS FOR CONTROLLING DIRECT-CYCLE NEUTRONIC REACTORS

DOEpatents

Reed, G.A.

1961-01-10

A control arrangement is offered for a boiling-water reactor. Boric acid is maintained in the water in the reactor and the amount in the reactor is controlled by continuously removing a portion of the water from the reactor, concentrating the boric acid by evaporating the water therefrom, returning a controlled amount of the acid to the reactor, and simultaneously controlling the water level by varying the rate of spent steam return to the reactor.

METHOD OF OPERATING A HEAVY WATER MODERATED REACTOR

DOEpatents

Vernon, H.C.

1962-08-14

A method of removing fission products from the heavy water used in a slurry type nuclear reactor is described. According to the process the slurry is steam distilled with carbon tetrachloride so that at least a part of the heavy water and carbon tetrachloride are vaporized; the heavy water and carbon tetrachloride are separated; the carbon tetrachloride is returned to the steam distillation column at different points in the column to aid in depositing the slurry particles at the bottom of the column; and the heavy water portion of the condensate is purified. (AEC)

Design of conduction cooling system for a high current HTS DC reactor

NASA Astrophysics Data System (ADS)

Dao, Van Quan; Kim, Taekue; Le Tat, Thang; Sung, Haejin; Choi, Jongho; Kim, Kwangmin; Hwang, Chul-Sang; Park, Minwon; Yu, In-Keun

2017-07-01

A DC reactor using a high temperature superconducting (HTS) magnet reduces the reactor’s size, weight, flux leakage, and electrical losses. An HTS magnet needs cryogenic cooling to achieve and maintain its superconducting state. There are two methods for doing this: one is pool boiling and the other is conduction cooling. The conduction cooling method is more effective than the pool boiling method in terms of smaller size and lighter weight. This paper discusses a design of conduction cooling system for a high current, high temperature superconducting DC reactor. Dimensions of the conduction cooling system parts including HTS magnets, bobbin structures, current leads, support bars, and thermal exchangers were calculated and drawn using a 3D CAD program. A finite element method model was built for determining the optimal design parameters and analyzing the thermo-mechanical characteristics. The operating current and inductance of the reactor magnet were 1,500 A, 400 mH, respectively. The thermal load of the HTS DC reactor was analyzed for determining the cooling capacity of the cryo-cooler. The study results can be effectively utilized for the design and fabrication of a commercial HTS DC reactor.

Effect of soaking, boiling, and steaming on total phenolic contentand antioxidant activities of cool season food legumes.

PubMed

Xu, Baojun; Chang, Sam K C

2008-09-01

The effects of soaking, boiling and steaming processes on the total phenolic components and antioxidant activity in commonly consumed cool season food legumes (CSFL's), including green pea, yellow pea, chickpea and lentil were investigated. As compared to original unprocessed legumes, all processing steps caused significant (p<0.05) decreases in total phenolic content (TPC), DPPH free radical scavenging activity (DPPH) in all tested CSFL's. All soaking and atmospheric boiling treatments caused significant (p<0.05) decreases in oxygen radical absorbing capacity (ORAC). However, pressure boiling and pressure steaming caused significant (p<0.05) increases in ORAC values. Steaming treatments resulted in a greater retention of TPC, DPPH, and ORAC values in all tested CSFL's as compared to boiling treatments. To obtain cooked legumes with similar palatability and firmness, pressure boiling shortened processing time as compared to atmospheric boiling, resulted in insignificant differences in TPC, DPPH for green and yellow pea. However, TPC and DPPH in cooked lentils differed significantly between atmospheric and pressure boiling. As compared to atmospheric processes, pressure processes significantly increased ORAC values in both boiled and steamed CSFL's. Greater TPC, DPPH and ORAC values were detected in boiling water than that in soaking and steaming water. Boiling also caused more solid loss than steaming. Steam processing exhibited several advantages in retaining the integrity of the legume appearance and texture of the cooked product, shortening process time, and greater retention of antioxidant components and activities. Copyright © 2008 Elsevier Ltd. All rights reserved.

Improved vortex reactor system

DOEpatents

Diebold, J.P.; Scahill, J.W.

1995-05-09

An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

Testimony of Fred R. Mynatt before the Energy Research and Development Subcommittee of the Committee on Science, Space, and Technology, US House of Representatives. [Advanced fuel technology, gas-cooled reactor technology, and liquid metal-cooled reactor technology programs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mynatt, F.R.

1987-03-18

This report provides a description of the statements submitted for the record to the committee on Science, Space, and Technology of the United States House of Representatives. These statements describe three principal areas of activity of the Advanced Reactor Technology Program of the Department of Energy (DOE). These areas are advanced fuel cycle technology, modular high-temperature gas-cooled reactor technology, and liquid metal-cooled reactor. The areas of automated reactor control systems, robotics, materials and structural design shielding and international cooperation were included in these statements describing the Oak Ridge National Laboratory's efforts in these areas. (FI)

Development of heat transfer enhancement techniques for external cooling of an advanced reactor vessel

NASA Astrophysics Data System (ADS)

Yang, Jun

Nucleate boiling is a well-recognized means for passively removing high heat loads (up to ˜106 W/m2) generated by a molten reactor core under severe accident conditions while maintaining relatively low reactor vessel temperature (<800 °C). With the upgrade and development of advanced power reactors, however, enhancing the nucleate boiling rate and its upper limit, Critical Heat Flux (CHF), becomes the key to the success of external passive cooling of reactor vessel undergoing core disrupture accidents. In the present study, two boiling heat transfer enhancement methods have been proposed, experimentally investigated and theoretically modelled. The first method involves the use of a suitable surface coating to enhance downward-facing boiling rate and CHF limit so as to substantially increase the possibility of reactor vessel surviving high thermal load attack. The second method involves the use of an enhanced vessel/insulation design to facilitate the process of steam venting through the annular channel formed between the reactor vessel and the insulation structure, which in turn would further enhance both the boiling rate and CHF limit. Among the various available surface coating techniques, metallic micro-porous layer surface coating has been identified as an appropriate coating material for use in External Reactor Vessel Cooling (ERVC) based on the overall consideration of enhanced performance, durability, the ease of manufacturing and application. Since no previous research work had explored the feasibility of applying such a metallic micro-porous layer surface coating on a large, downward facing and curved surface such as the bottom head of a reactor vessel, a series of characterization tests and experiments were performed in the present study to determine a suitable coating material composition and application method. Using the optimized metallic micro-porous surface coatings, quenching and steady-state boiling experiments were conducted in the Sub-scale Boundary Layer Boiling (SBLB) test facility at Penn State to investigate the nucleate boiling and CHF enhancement effects of the surface coatings by comparing the measurements with those for a plain vessel without coatings. An overall enhancement in nucleate boiling rates and CHF limits up to 100% were observed. Moreover, combination of data from quenching experiments and steady-state experiments produced new sets of boiling curves, which covered both the nucleate and transient boiling regimes with much greater accuracy. Beside the experimental work, a theoretical CHF model has also been developed by considering the vapor dynamics and the boiling-induced two-phase motions in three separate regions adjacent to the heating surface. The CHF model is capable of predicting the performance of micro-porous coatings with given particle diameter, porosity, media permeability and thickness. It is found that the present CHF model agrees favorably with the experimental data. Effects of an enhanced vessel/insulation structure on the local nucleate boiling rate and CHF limit have also been investigated experimentally. It is observed that the local two-phase flow quantities such as the local void fraction, quality, mean vapor velocity, mean liquid velocity, and mean vapor and liquid mass flow rates could have great impact on the local surface heat flux as boiling of water takes place on the vessel surface. An upward co-current two-phase flow model has been developed to predict the local two-phase flow behavior for different flow channel geometries, which are set by the design of insulation structures. It is found from the two-phase flow visualization experiments and the two-phase flow model calculations that the enhanced vessel/insulation structure greatly improved the steam venting process at the minimum gap location compared to the performance of thermal insulation structures without enhancement. Moveover, depending on the angular location, steady-state boiling experiments with the enhanced insulation design showed an enhancement of 1.8 to 3.0 times in the local critical heat flux. Finally, nucleate boiling and CHF correlations were developed based on the data obtained from various quenching and steady-state boiling experiments. Additionally, CHF enhancement factors were determined and examined to show the separate and integral effects of the two ERVC enhancement methods. When both vessel coating and insulation structure were used simultaneously, the integral effect on CHF enhancement was found much less than the product of the two separate effects, indicating possible competing mechanisms (i.e., interference) between the two enhancement methods.

Stabilization of Hydrogen Production via Methanol Steam Reforming in Microreactor by Al2O3 Nano-Film Enhanced Catalyst Adhesion.

PubMed

Jeong, Heondo; Na, Jeong-Geol; Jang, Min Su; Ko, Chang Hyun

2016-05-01

In hydrogen production by methanol steam reforming reaction with microchannel reactor, Al2O3 thin film formed by atomic layer deposition (ALD) was introduced on the surface of microchannel reactor prior to the coating of catalyst particles. Methanol conversion rate and hydrogen production rate, increased in the presence of Al2O3 thin film. Over-view and cross-sectional scanning electron microscopy study showed that the adhesion between catalyst particles and the surface of microchannel reactor enhanced due to the presence of Al2O3 thin film. The improvement of hydrogen production rate inside the channels of microreactor mainly came from the stable fixation of catalyst particles on the surface of microchannels.

Computer model of catalytic combustion/Stirling engine heater head

NASA Technical Reports Server (NTRS)

Chu, E. K.; Chang, R. L.; Tong, H.

1981-01-01

The basic Acurex HET code was modified to analyze specific problems for Stirling engine heater head applications. Specifically, the code can model: an adiabatic catalytic monolith reactor, an externally cooled catalytic cylindrical reactor/flat plate reactor, a coannular tube radiatively cooled reactor, and a monolithic reactor radiating to upstream and downstream heat exchangers.

Investigation of reactions in a fluidized bed reactor during chemical looping combustion of coal/steam with copper oxide-iron oxide-alumina oxygen carrier

DOE Office of Scientific and Technical Information (OSTI.GOV)

Siriwardane, Ranjani; Benincosa, William; Riley, Jarrett

This paper presents data on conversion of two different coals with a chemical looping oxygen carrier, CuO-Fe 2O 3-alumina, and over a range of conditions including steam and various levels of reduction of the oxygen carrier. Reactions of coal/steam/CuO-Fe 2O 3-alumina oxygen carrier and coal/steam/partially reduced CuO-Fe 2O 3-alumina oxygen carrier were investigated with Wyodak coal and Illinois #6 coal in a fluidized bed reactor. Temperature programmed reaction studies indicated that the oxygen carrier enhanced the steam gasification/combustion rates of both coals. Rates of gasification/combustion were higher with Wyodak coal (sub bituminous) than that with Illinois #6 coal (bituminous). Inmore » addition to the increase in reaction rates, the total moles of carbon that were gasified and combusted from coal/steam increased in the presence of the oxygen carrier. The reduced oxygen carrier promoted the water-gas shift reaction when reacted with synthesis gas in the presence of steam, but the reverse water gas shift reaction was observed when steam was not present. The partially reduced oxygen carrier enhanced the production of H 2 from coal/steam, which was different from the observations with un-reduced oxygen carrier. Water splitting reaction to produce H 2 was also observed with the reduced oxygen carrier. CuO-Fe 2O 3-alumina reacted with coal during the temperature ramp to 850 °C even in the absence of steam due to the chemical-looping oxygen uncoupling (CLOU) reaction. Here, the fourier transform infra-red (FTIR) analysis indicated the presence of volatile aromatics during the temperature ramp and these may have also contributed to the reactions with the oxygen carrier in the absence of steam. Increasing steam concentration had a negative effect on the CLOU reaction.« less

Investigation of reactions in a fluidized bed reactor during chemical looping combustion of coal/steam with copper oxide-iron oxide-alumina oxygen carrier

DOE PAGES

Siriwardane, Ranjani; Benincosa, William; Riley, Jarrett; ...

2016-10-06

This paper presents data on conversion of two different coals with a chemical looping oxygen carrier, CuO-Fe 2O 3-alumina, and over a range of conditions including steam and various levels of reduction of the oxygen carrier. Reactions of coal/steam/CuO-Fe 2O 3-alumina oxygen carrier and coal/steam/partially reduced CuO-Fe 2O 3-alumina oxygen carrier were investigated with Wyodak coal and Illinois #6 coal in a fluidized bed reactor. Temperature programmed reaction studies indicated that the oxygen carrier enhanced the steam gasification/combustion rates of both coals. Rates of gasification/combustion were higher with Wyodak coal (sub bituminous) than that with Illinois #6 coal (bituminous). Inmore » addition to the increase in reaction rates, the total moles of carbon that were gasified and combusted from coal/steam increased in the presence of the oxygen carrier. The reduced oxygen carrier promoted the water-gas shift reaction when reacted with synthesis gas in the presence of steam, but the reverse water gas shift reaction was observed when steam was not present. The partially reduced oxygen carrier enhanced the production of H 2 from coal/steam, which was different from the observations with un-reduced oxygen carrier. Water splitting reaction to produce H 2 was also observed with the reduced oxygen carrier. CuO-Fe 2O 3-alumina reacted with coal during the temperature ramp to 850 °C even in the absence of steam due to the chemical-looping oxygen uncoupling (CLOU) reaction. Here, the fourier transform infra-red (FTIR) analysis indicated the presence of volatile aromatics during the temperature ramp and these may have also contributed to the reactions with the oxygen carrier in the absence of steam. Increasing steam concentration had a negative effect on the CLOU reaction.« less

Nuclear reactor building

DOEpatents

Gou, P.F.; Townsend, H.E.; Barbanti, G.

1994-04-05

A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

Nuclear reactor building

DOEpatents

Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

1994-01-01

A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

System Analysis for Decay Heat Removal in Lead-Bismuth Cooled Natural Circulated Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takaaki Sakai; Yasuhiro Enuma; Takashi Iwasaki

2002-07-01

Decay heat removal analyses for lead-bismuth cooled natural circulation reactors are described in this paper. A combined multi-dimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural circulation reactors. For the preliminary study, transient analysis has been performed for a 100 MWe lead-bismuth-cooled reactor designed by Argonne National Laboratory (ANL). In addition, decay heat removal characteristics of a 400 MWe lead-bismuth-cooled natural circulation reactor designed by Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. PRACS (Primary Reactor Auxiliary Cooling System) is prepared for the JNC's concept to get sufficient heatmore » removal capacity. During 2000 sec after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 Centigrade, because the buoyancy force in a primary circulation path is temporary reduced. However, the natural circulation is recovered by the PRACS system and the out let temperature decreases successfully. (authors)« less

System Analysis for Decay Heat Removal in Lead-Bismuth-Cooled Natural-Circulation Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sakai, Takaaki; Enuma, Yasuhiro; Iwasaki, Takashi

2004-03-15

Decay heat removal analyses for lead-bismuth-cooled natural-circulation reactors are described in this paper. A combined multidimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural-circulation reactors. For the preliminary study, transient analysis has been performed for a 300-MW(thermal) lead-bismuth-cooled reactor designed by Argonne National Laboratory. In addition, decay heat removal characteristics of a 400-MW(electric) lead-bismuth-cooled natural-circulation reactor designed by the Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. The primary reactor auxiliary cooling system (PRACS) is prepared for the JNC concept to get sufficient heat removal capacity. During 2000 smore » after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 deg. C because the buoyancy force in a primary circulation path is temporarily reduced. However, the natural circulation is recovered by the PRACS system, and the outlet temperature decreases successfully.« less

Control of reactor coolant flow path during reactor decay heat removal

DOEpatents

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Multidimensional Mixing Behavior of Steam-Water Flow in a Downcomer Annulus During LBLOCA Reflood Phase with a Direct Vessel Injection Mode

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kwon, Tae-Soon; Yun, Byong-Jo; Euh, Dong-Jin

Multidimensional thermal-hydraulic behavior in the downcomer annulus of a pressurized water reactor (PWR) vessel with a direct vessel injection mode is presented based on the experimental observation in the MIDAS (multidimensional investigation in downcomer annulus simulation) steam-water test facility. From the steady-state test results to simulate the late reflood phase of a large-break loss-of-coolant accident (LBLOCA), isothermal lines show the multidimensional phenomena of a phasic interaction between steam and water in the downcomer annulus very well. MIDAS is a steam-water separate effect test facility, which is 1/4.93 linearly scaled down to a 1400-MW(electric) PWR type of a nuclear reactor, focusedmore » on understanding multidimensional thermal-hydraulic phenomena in a downcomer annulus with various types of safety injection during the refill or reflood phase of an LBLOCA. The initial and the boundary conditions are scaled from the pretest analysis based on the preliminary calculation using the TRAC code. The superheated steam with a superheating degree of 80 K at a given downcomer pressure of 180 kPa is injected equally through three intact cold legs into the downcomer.« less

Production of synthetic fuels using syngas from a steam hydrogasification and reforming process

NASA Astrophysics Data System (ADS)

Raju, Arun Satheesh Kumar

This thesis is aimed at the research, optimization and development of a thermo-chemical process aimed at the production of synthesis gas (mixture of H2 and CO) with a flexible H2 to CO ratio using coupled steam hydrogasification and steam reforming processes. The steam hydrogasification step generates a product gas containing significant amounts of methane by gasifying a carbonaceous feed material with steam and internally generated H2. This product gas is converted to synthesis gas with an excess H2 to CO using the steam reformer. Research involving experimental and simulation work has been conducted on steam hydrogasification, steam reforming and the Fischer-Tropsch reaction. The Aspen Plus simulation tool has been used to develop a process model that can perform heat and mass balance calculations of the whole process using built-in reactor modules and an empirical FT model available in the literature. This model has been used to estimate optimum feed ratios and process conditions for specific feedstocks and products. Steam hydrogasification of coal and wood mixtures of varying coal to wood ratios has been performed in a stirred batch reactor. The carbon conversion of the feedstocks to gaseous products is around 60% at 700°C and 80% at 800°C. The coal to wood ratio of the feedstock does not exert a significant influence on the carbon conversion. The rates of formation of CO, CO 2 and CH4 during gasification have been calculated based on the experimental results using a simple kinetic model. Experimental research on steam reforming has been performed. It has been shown that temperature and the feed CO2/CH4 ratio play a dominant role in determining the product gas H2/CO ratio. Reforming of typical steam hydrogasification product-gas stream has been investigated over a commercial steam reforming catalyst. The results demonstrate that the combined use of steam hydrogasification process with a reformer can generate a synthesis gas with a predetermined H2/CO ratio from carbonaceous feedstocks. Experimental work on the Fischer-Tropsch synthesis has also been performed. A life cycle analysis has been performed with the objective of comparing the life cycle energy consumption and emissions of synthetic diesel fuel produced through the CE-CERT process with other fuel/vehicle combinations. The experimental and simulation results presented here demonstrate that the CE-CERT process is versatile and can potentially handle a number of different feedstocks. CE-CERT process appears to be suitable for commercialization in very large scales with a coal feedstock and also in a distributed network of smaller scale reactors utilizing localized renewable feedstocks.

Westinghouse Small Modular Reactor passive safety system response to postulated events

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, M. C.; Wright, R. F.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. Themore » integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000{sup R} reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to reduce the reactor pressure in a controlled manner to facilitate the passive injection. Long-term decay heat removal is accomplished using the passive heat removal systems augmented by heat transfer through the containment vessel to the environment. The passive injection systems are designed so that the fuel remains covered and effectively cooled throughout the event. Like during the frequent faults, the passive systems provide effective cooling without the need for ac power for seven days following the accident. Connections are available to add additional water to indefinitely cool the plant. The response of the safety systems of the Westinghouse SMR to various initiating faults has been examined. Among them, two accidents; an extended station blackout event, and a LOCA event have been evaluated to demonstrate how the plant will remain safe in the unlikely event that either should occur. (authors)« less

Analysis of boron dilution in a four-loop PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sun, J.G.; Sha, W.T.

1995-03-01

Thermal mixing and boron dilution in a pressurized water reactor were analyzed with COMMIX codes. The reactor system was the four-loop Zion reactor. Two boron dilution scenarios were analyzed. In the first scenario, the plant is in cold shutdown and the reactor coolant system has just been filled after maintenance on the steam generators. To flush the air out of the steam generator tubes, a reactor coolant pump (RCP) is started, with the water in the pump suction line devoid of boron and at the same temperature as the coolant in the system. In the second scenario, the plant ismore » at hot standby and the reactor coolant system has been heated to operating temperature after a long outage. It is assumed that an RCP is started, with the pump suction line filled with cold unborated water, forcing a slug of diluted coolant down the downcomer and subsequently through the reactor core. The subsequent transient thermal mixing and boron dilution that would occur in the reactor system is simulated for these two scenarios. The reactivity insertion rate and the total reactivity are evaluated and a sensitivity study is performed to assess the accuracy of the numerical modeling of the geometry of the reactor coolant system.« less

Performance calculations for 200-1000 MWe MHD/steam power plants

NASA Technical Reports Server (NTRS)

Staiger, P. J.

1981-01-01

The effects of MHD generator length, level of oxygen enrichment, and oxygen production power on the performance of MHD/steam power plants ranging from 200 to 1000 MW in electrical output are investigated. The plants considered use oxygen enriched combustion air preheated to 1100 F. Both plants in which the MHD generator is cooled with low temperature and pressure boiler feedwater and plants in which the generator is cooled with high temperature and pressure boiler feedwater are considered. For plants using low temperature boiler feedwater for generator cooling the maximum thermodynamic efficiency is obtained with shorter generators and a lower level of oxygen enrichment compared to plants using high temperature boiler feedwater for generator cooling. The generator length at which the maximum plant efficiency occurs increases with power plant size for plants with a generator cooled by low temperature feedwater. Also shown is the relationship of the magnet stored energy requirement of the generator length and the power plant performance. Possible cost/performance tradeoffs between magnet cost and plant performance are indicated.

Simulation of heat and mass transfer processes in the experimental section of the air-condensing unit of Scientific Production Company "Turbocon"

NASA Astrophysics Data System (ADS)

Artemov, V. I.; Minko, K. B.; Yan'kov, G. G.; Kiryukhin, A. V.

2016-05-01

A mathematical model was developed to be used for numerical analysis of heat and mass transfer processes in the experimental section of the air condenser (ESAC) created in the Scientific Production Company (SPC) "Turbocon" and mounted on the territory of the All-Russia Thermal Engineering Institute. The simulations were performed using the author's CFD code ANES. The verification of the models was carried out involving the experimental data obtained in the tests of ESAC. The operational capability of the proposed models to calculate the processes in steam-air mixture and cooling air and algorithms to take into account the maldistribution in the various rows of tube bundle was shown. Data on the influence of temperature and flow rate of the cooling air on the pressure in the upper header of ESAC, effective heat transfer coefficient, steam flow distribution by tube rows, and the dimensions of the ineffectively operating zones of tube bundle for two schemes of steam-air mixture flow (one-pass and two-pass ones) were presented. It was shown that the pressure behind the turbine (in the upper header) increases significantly at increase of the steam flow rate and reduction of the flow rate of cooling air and its temperature rise, and the maximum value of heat transfer coefficient is fully determined by the flow rate of cooling air. Furthermore, the steam flow rate corresponding to the maximum value of heat transfer coefficient substantially depends on the ambient temperature. The analysis of the effectiveness of the considered schemes of internal coolant flow was carried out, which showed that the two-pass scheme is more effective because it provides lower pressure in the upper header, despite the fact that its hydraulic resistance at fixed flow rate of steam-air mixture is considerably higher than at using the one-pass schema. This result is a consequence of the fact that, in the two-pass scheme, the condensation process involves the larger internal surface of tubes, results in lower values of Δ t (the temperature difference between internal and external coolant) for a given heat load.

Hydrogen Production by Sorption Enhanced Steam Reforming (SESR) of Biomass in a Fluidised-Bed Reactor Using Combined Multifunctional Particles

PubMed Central

Zheng, Liya; Zhang, Zili; Fennell, Paul S.

2018-01-01

The performance of combined CO2-sorbent/catalyst particles for sorption enhanced steam reforming (SESR), prepared via a simple mechanical mixing protocol, was studied using a spout-fluidised bed reactor capable of continuous solid fuel (biomass) feeding. The influence of particle size (300–500 and 710–1000 µm), CaO loading (60–100 wt %), Ni-loading (10–40 wt %) and presence of dicalcium silicate support (22.6 wt %) on SESR process performance were investigated. The combined particles were characterised by their density, porosity and CO2 carrying capacity with the analysis by thermogravimetric analysis (TGA), Brunauer-Emmett-Teller (BET), Barrett-Joyner-Halenda (BJH) and mercury intrusion porosimetry (MIP). All experiments were conducted with continuous oak biomass feeding at a rate of 0.9 g/min ± 10%, and the reactor was operated at 660 ± 5 °C, 1 atm and 20 ± 2 vol % steam which corresponds to a steam-to-carbon ratio of 1.2:1. Unsupported combined particles containing 21.0 wt % Ni and 79 wt % CaO were the best performing sorbent/catalyst particle screened in this study, when accounting for the cost of Ni and the improvement in H2 produced by high Ni content particles. SESR tests with these combined particles produced 61 mmol H2/gbiomass (122 g H2/kgbiomass) at a purity of 61 vol %. Significant coke formation within the feeding tube and on the surfaces of the particles was observed which was attributed to the low steam to carbon ratio utilised. PMID:29883427

77 FR 63343 - Biweekly Notice: Applications and Amendments to Facility Operating Licenses and Combined Licenses...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-16

... PWR [Pressurized-Water Reactor] Steam Generator Tubes'' (Reference 32) and [Nuclear Energy Institute... maintains the required structural margins of the SG tubes for both normal and accident conditions. Nuclear Energy Institute 97-06, ``Steam Generator Program Guidelines'' (Reference 8), and NRC Regulatory Guide 1...

Downhole steam generator having a downhole oxidant compressor

DOEpatents

Fox, R.L.

1981-01-07

Am improved apparatus is described for the downhole injection of steam into boreholes, for tertiary oil recovery. It includes an oxidant supply, a fuel supply, an igniter, a water supply, an oxidant compressor, and a combustor assembly. The apparatus is designed for efficiency, preheating of the water, and cooling of the combustion chamber walls. The steam outlet to the borehole is provided with pressure-responsive doors for closing the outlet in response to flameout. (DLC)

METHOD OF FIXING NITROGEN FOR PRODUCING OXIDES OF NITROGEN

DOEpatents

Harteck, P.; Dondes, S.

1959-08-01

A method is described for fixing nitrogen from air by compressing the air, irradiating the compressed air in a nuclear reactor, cooling to remove NO/ sub 2/, compressing the cooled gas, further cooling to remove N/sub 2/O and recirculating the cooled compressed air to the reactor.

Horizontal steam generator thermal-hydraulics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ubra, O.; Doubek, M.

1995-09-01

Horizontal steam generators are typical components of nuclear power plants with pressure water reactor type VVER. Thermal-hydraulic behavior of horizontal steam generators is very different from the vertical U-tube steam generator, which has been extensively studied for several years. To contribute to the understanding of the horizontal steam generator thermal-hydraulics a computer program for 3-D steady state analysis of the PGV-1000 steam generator has been developed. By means of this computer program, a detailed thermal-hydraulic and thermodynamic study of the horizontal steam generator PGV-1000 has been carried out and a set of important steam generator characteristics has been obtained. Themore » 3-D distribution of the void fraction and 3-D level profile as functions of load and secondary side pressure have been investigated and secondary side volumes and masses as functions of load and pressure have been evaluated. Some of the interesting results of calculations are presented in the paper.« less

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.

Controls on Water Use for Thermoelectric Generation: Case Study Texas, U.S.

PubMed Central

2013-01-01

Large-scale U.S. dependence on thermoelectric (steam electric) generation requiring water for cooling underscores the need to understand controls on this water use. The study objective was to quantify water consumption and withdrawal for thermoelectric generation, identifying controls, using Texas as a case study. Water consumption for thermoelectricity in Texas in 2010 totaled ∼0.43 million acre feet (maf; 0.53 km3), accounting for ∼4% of total state water consumption. High water withdrawals (26.2 maf, 32.3 km3) mostly reflect circulation between ponds and power plants, with only two-thirds of this water required for cooling. Controls on water consumption include (1) generator technology/thermal efficiency and (2) cooling system, resulting in statewide consumption intensity for natural gas combined cycle generators with mostly cooling towers (0.19 gal/kWh) being 63% lower than that of traditional coal, nuclear, or natural gas steam turbine generators with mostly cooling ponds (0.52 gal/kWh). The primary control on water withdrawals is cooling system, with ∼2 orders of magnitude lower withdrawals for cooling towers relative to once-through ponds statewide. Increases in natural gas combined cycle plants with cooling towers in response to high production of low-cost natural gas has greatly reduced water demand for thermoelectric cooling since 2000. PMID:23937226

Controls on water use for thermoelectric generation: case study Texas, US.

PubMed

Scanlon, Bridget R; Reedy, Robert C; Duncan, Ian; Mullican, William F; Young, Michael

2013-10-01

Large-scale U.S. dependence on thermoelectric (steam electric) generation requiring water for cooling underscores the need to understand controls on this water use. The study objective was to quantify water consumption and withdrawal for thermoelectric generation, identifying controls, using Texas as a case study. Water consumption for thermoelectricity in Texas in 2010 totaled ∼0.43 million acre feet (maf; 0.53 km(3)), accounting for ∼4% of total state water consumption. High water withdrawals (26.2 maf, 32.3 km(3)) mostly reflect circulation between ponds and power plants, with only two-thirds of this water required for cooling. Controls on water consumption include (1) generator technology/thermal efficiency and (2) cooling system, resulting in statewide consumption intensity for natural gas combined cycle generators with mostly cooling towers (0.19 gal/kWh) being 63% lower than that of traditional coal, nuclear, or natural gas steam turbine generators with mostly cooling ponds (0.52 gal/kWh). The primary control on water withdrawals is cooling system, with ∼2 orders of magnitude lower withdrawals for cooling towers relative to once-through ponds statewide. Increases in natural gas combined cycle plants with cooling towers in response to high production of low-cost natural gas has greatly reduced water demand for thermoelectric cooling since 2000.

Cooling systems for ultra-high temperature turbines.

PubMed

Yoshida, T

2001-05-01

This paper describes an introduction of research and development activities on steam cooling in gas turbines at elevated temperature of 1500 C and 1700 C level, partially including those on water cooling. Descriptions of a new cooling system that employs heat pipes are also made. From the view point of heat transfer, its promising applicability is shown with experimental data and engine performance numerical evaluation.

Modification of wood fiber using steam

Treesearch

Roger Rowell; Sandra Lange; Jim McSweeny; Mark Davis

2002-01-01

High temperature steam treatment of wood fiber in a closed press during fiberboard pressing and then cooling the fiberboard while still under pressure to below the glass transition temperature of lignin, greatly increased the dimensional stability and decreased the hemicellulose content of the fiberboards produced. For example, after pressing aspen fiber four minutes...

Thermodynamic analysis of the advanced zero emission power plant

NASA Astrophysics Data System (ADS)

Kotowicz, Janusz; Job, Marcin

2016-03-01

The paper presents the structure and parameters of advanced zero emission power plant (AZEP). This concept is based on the replacement of the combustion chamber in a gas turbine by the membrane reactor. The reactor has three basic functions: (i) oxygen separation from the air through the membrane, (ii) combustion of the fuel, and (iii) heat transfer to heat the oxygen-depleted air. In the discussed unit hot depleted air is expanded in a turbine and further feeds a bottoming steam cycle (BSC) through the main heat recovery steam generator (HRSG). Flue gas leaving the membrane reactor feeds the second HRSG. The flue gas consist mainly of CO2 and water vapor, thus, CO2 separation involves only the flue gas drying. Results of the thermodynamic analysis of described power plant are presented.

Experimental research of heterogeneous nuclei in superheated steam

NASA Astrophysics Data System (ADS)

Bartoš, Ondřej; Kolovratník, Michal; Šmíd, Bohuslav; Hrubý, Jan

2016-03-01

A mobile steam expansion chamber has been developed to investigate experimentally homogeneous and heterogeneous nucleation processes in steam, both in the laboratory and at power plants using the steam withdrawn from the steam turbine. The purpose of the device is to provide new insight into the physics of nonequilibrium wet steam formation, which is one of the factors limiting the efficiency and reliability of steam turbines. The expanded steam or a mixture of steam with a non-condensable gas rapidly expands in the expansion chamber. Due to adiabatic cooling, the temperature drops below the dew point of the steam at a given pressure. When reaching a sufficiently high supersaturation, droplets are nucleated. By tuning the supersaturation in the so-called nucleation pulse, particles of various size ranges can be activated. This fact is used in the present study to measure the aerosol particles present in the air. Homogeneous nucleation was negligible in this case. The experiment demonstrates the functionality of the device, data acquisition system and data evaluation methods.

Thermodynamic analysis of the efficiency of high-temperature steam electrolysis system for hydrogen production

NASA Astrophysics Data System (ADS)

Mingyi, Liu; Bo, Yu; Jingming, Xu; Jing, Chen

High-temperature steam electrolysis (HTSE), a reversible process of solid oxide fuel cell (SOFC) in principle, is a promising method for highly efficient large-scale hydrogen production. In our study, the overall efficiency of the HTSE system was calculated through electrochemical and thermodynamic analysis. A thermodynamic model in regards to the efficiency of the HTSE system was established and the quantitative effects of three key parameters, electrical efficiency (η el), electrolysis efficiency (η es), and thermal efficiency (η th) on the overall efficiency (η overall) of the HTSE system were investigated. Results showed that the contribution of η el, η es, η th to the overall efficiency were about 70%, 22%, and 8%, respectively. As temperatures increased from 500 °C to 1000 °C, the effect of η el on η overall decreased gradually and the η es effect remained almost constant, while the η th effect increased gradually. The overall efficiency of the high-temperature gas-cooled reactor (HTGR) coupled with the HTSE system under different conditions was also calculated. With the increase of electrical, electrolysis, and thermal efficiency, the overall efficiencies were anticipated to increase from 33% to a maximum of 59% at 1000 °C, which is over two times higher than that of the conventional alkaline water electrolysis.

Grain growth in uranium nitride prepared by spark plasma sintering

NASA Astrophysics Data System (ADS)

Johnson, Kyle D.; Lopes, Denise Adorno

2018-05-01

Uranium mononitride (UN) has long been considered a potential high density, high performance fuel candidate for light water reactor (LWR) and fast reactor (FR) applications. However, deployability of this fuel has been limited by the notable resistance to sintering and subsequent difficulty in producing a desirable microstructure, the high costs associated with 15N enrichment, as well as the known proclivity to oxidation and interaction with steam. In this study, the stimulation of grain growth in UN pellets sintered using SPS has been investigated. The results reveal that by using SPS and controlling temperature, time, and holding pressure, grain growth can be stimulated and controlled to produce a material featuring both a desired porosity and grain size, at least within the range of interest for nuclear fuel candidates. Grain sizes up to 31 μm were obtained using temperatures of 1650 °C and hold times of 15 min. Evaluation by EBSD reveal grain rotation and coalescence as the dominant mechanism in grain growth, which is suppressed by the application of higher external pressure. Moreover, complete closure of the porosity of the material was observed at relative densities of 96% TD, resulting in a material with sufficient porosity to accommodate LWR burnup. These results indicate that a method exists for the economic fabrication of an 15N-bearing uranium mononitride fuel with favorable microstructural characteristics compatible with use in a light water-cooled nuclear reactor.

Posttest data analysis of FIST experimental TRAC-BD1/MOD1 power transient experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wheatley, P.D.; Wagner, K.C.

The FIST power transient test 6PMC2 was analyzed to further the understanding of the FIST facility and provide an assessment of TRAC-BD1/MOD1. FIST power transient 6PMC2 investigated the thermal-hydraulic response following inadvertent closure of the main steam isolation valve and the subsequent failure of the reactor to scram. Failure of the high pressure core spray system was also assumed, resulting in only the reactor core isolation cooling flow for inventory makeup during the transient. The experiment was a sensitivity study with relatively high core power and low makeup rates. This study provides one of the first opportunities to assess TRAC-BD1/MOD1more » under power transient and natural circulation conditions with data from a facility with prototypical BWR geometry. The power transient test was analyzed with emphasis on the following phenomena: (a) the system pressure response, (b) the natural circulation flows and rates, and (c) the heater rod cladding temperature response. Based on the results of this study, TRAC-BD1/MOD1 can be expected to calculate the thermal-hydraulic behavior of a BWR during a power transient.« less

Filmwise Condensation of Steam on Externally-Finned Horizontal Tubes.

DTIC Science & Technology

1983-12-01

via gravity to the boiler. The auxiliary condenser was constructed of two 9.5-mm (3/8-in) water- cooled ccpper lines helically coiled to a height of...34. " . .. . ’ . .- .. ’. .. .- . . . i . ’ -, - NPS69-83-003 - m NAVAL POSTGRADUATE SCHOOL Monterey, California THESIS FILMWISE CONDENSATION OF STEAM ON EXTERNALLY-FINNED...and SubEtee) 5. TYPE OF REPORT & PERIOD COVERED Filmwise Condensation of Steam on Master Thesis; Externally-Finned Horizontal Tubes D e r1 6

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)

77 FR 47123 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-07

...-Water Reactors] Steam Generator Tubes'' [Reference 7] and NEI [Nuclear Energy Institute] 97-06, ``Steam... CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these... the required structural margins of the SG tubes for both normal and accident conditions. Nuclear...

Start-up control system and vessel for LMFBR

DOEpatents

Durrant, Oliver W.; Kakarala, Chandrasekhara R.; Mandel, Sheldon W.

1987-01-01

A reflux condensing start-up system includes a steam generator, a start-up vessel connected parallel to the steam generator, a main steam line connecting steam outlets of the steam generator and start-up vessel to a steam turbine, a condenser connected to an outlet of the turbine and a feedwater return line connected between the condenser and inlets of the steam generator and start-up vessel. The start-up vessel has one or more heaters at the bottom thereof for heating feedwater which is supplied over a start-up line to the start-up vessel. Steam is thus generated to pressurize the steam generator before the steam generator is supplied with a heat transfer medium, for example liquid sodium, in the case of a liquid metal fast breeder reactor. The start-up vessel includes upper and lower bulbs with a smaller diameter mid-section to act as water and steam reservoirs. The start-up vessel can thus be used not only in a start-up operation but as a mixing tank, a water storage tank and a level control at low loads for controlling feedwater flow.

Start-up control system and vessel for LMFBR

DOEpatents

Durrant, Oliver W.; Kakarala, Chandrasekhara R.; Mandel, Sheldon W.

1987-01-01

A reflux condensing start-up system comprises a steam generator, a start-up vessel connected parallel to the steam generator, a main steam line connecting steam outlets of the steam generator and start-up vessel to a steam turbine, a condenser connected to an outlet of the turbine and a feedwater return line connected between the condenser and inlets of the steam generator and start-up vessel. The start-up vessel has one or more heaters at the bottom thereof for heating feedwater which is supplied over a start-up line to the start-up vessel. Steam is thus generated to pressurize the steam generator before the steam generator is supplied with a heat transfer medium, for example liquid sodium, in the case of a liquid metal fast breeder reactor. The start-up vessel includes upper and lower bulbs with a smaller diameter mid-section to act as water and steam reservoirs. The start-up vessel can thus be used not only in a start-up operation but as a mixing tank, a water storage tank and a level control at low loads for controlling feedwater flow.

Method and apparatus for enhancing reactor air-cooling system performance

DOEpatents

Hunsbedt, Anstein

1996-01-01

An enhanced decay heat removal system for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer.

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

1993-01-01

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

Experimental investigation of biomimetic self-pumping and self-adaptive transpiration cooling.

PubMed

Jiang, Pei-Xue; Huang, Gan; Zhu, Yinhai; Xu, Ruina; Liao, Zhiyuan; Lu, Taojie

2017-09-01

Transpiration cooling is an effective way to protect high heat flux walls. However, the pumps for the transpiration cooling system make the system more complex and increase the load, which is a huge challenge for practical applications. A biomimetic self-pumping transpiration cooling system was developed inspired by the process of trees transpiration that has no pumps. An experimental investigation showed that the water coolant automatically flowed from the water tank to the hot surface with a height difference of 80 mm without any pumps. A self-adaptive transpiration cooling system was then developed based on this mechanism. The system effectively cooled the hot surface with the surface temperature kept to about 373 K when the heating flame temperature was 1639 K and the heat flux was about 0.42 MW m -2 . The cooling efficiency reached 94.5%. The coolant mass flow rate adaptively increased with increasing flame heat flux from 0.24 MW m -2 to 0.42 MW m -2 while the cooled surface temperature stayed around 373 K. Schlieren pictures showed a protective steam layer on the hot surface which blocked the flame heat flux to the hot surface. The protective steam layer thickness also increased with increasing heat flux.

Biomass gasification for liquid fuel production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Najser, Jan, E-mail: jan.najser@vsb.cz, E-mail: vaclav.peer@vsb.cz; Peer, Václav, E-mail: jan.najser@vsb.cz, E-mail: vaclav.peer@vsb.cz; Vantuch, Martin

2014-08-06

In our old fix-bed autothermal gasifier we tested wood chips and wood pellets. We make experiments for Czech company producing agro pellets - pellets made from agricultural waste and fastrenewable natural resources. We tested pellets from wheat and rice straw and hay. These materials can be very perspective, because they dońt compete with food production, they were formed in sufficient quantity and in the place of their treatment. New installation is composed of allothermal biomass fixed bed gasifier with conditioning and using produced syngas for Fischer - Tropsch synthesis. As a gasifying agent will be used steam. Gas purification willmore » have two parts - separation of dust particles using a hot filter and dolomite reactor for decomposition of tars. In next steps, gas will be cooled, compressed and removed of sulphur and chlorine compounds and carbon dioxide. This syngas will be used for liquid fuel synthesis.« less

Biomass gasification for liquid fuel production

NASA Astrophysics Data System (ADS)

Najser, Jan; Peer, Václav; Vantuch, Martin

2014-08-01

In our old fix-bed autothermal gasifier we tested wood chips and wood pellets. We make experiments for Czech company producing agro pellets - pellets made from agricultural waste and fastrenewable natural resources. We tested pellets from wheat and rice straw and hay. These materials can be very perspective, because they dońt compete with food production, they were formed in sufficient quantity and in the place of their treatment. New installation is composed of allothermal biomass fixed bed gasifier with conditioning and using produced syngas for Fischer - Tropsch synthesis. As a gasifying agent will be used steam. Gas purification will have two parts - separation of dust particles using a hot filter and dolomite reactor for decomposition of tars. In next steps, gas will be cooled, compressed and removed of sulphur and chlorine compounds and carbon dioxide. This syngas will be used for liquid fuel synthesis.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

A multiphysics analysis system for sodium-water reaction phenomena in a steam generator of sodium-cooled fast reactors was newly developed. The analysis system consists of the mechanistic numerical analysis codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow and sodium-water chemical reaction caused by discharging of pressurized water vapor. Applicability of the SERAPHIM code was confirmed through the analyses of the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models integratedmore » into the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the tube rapidly heated by the reacting jet. The developed system enables evaluation of the wastage environment and the possibility of the failure propagation.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Llopis, C.; Mendizabal, R.; Perez, J.

An assessment of RELAP5/MOD2 cycle 36.04 against a load rejection from 100% to 50% power in Vandals II NPP (Spain) is presented. The work is inscribed in the framework of the Spanish contribution to ICAP Project. The model used in the simulation consists of a single loop, a steam generator and a steam line up to the steam header all of them enlarged on a scale of 3:1, and full-scaled reactor vessel and pressurizer. The results of the calculations have been in reasonable agreement with plant measurements.

Method for increasing steam decomposition in a coal gasification process

DOEpatents

Wilson, Marvin W.

1988-01-01

The gasification of coal in the presence of steam and oxygen is significantly enhanced by introducing a thermochemical water-splitting agent such as sulfuric acid, into the gasifier for decomposing the steam to provide additional oxygen and hydrogen usable in the gasification process for the combustion of the coal and enrichment of the gaseous gasification products. The addition of the water-splitting agent into the gasifier also allows for the operation of the reactor at a lower temperature.

Method for increasing steam decomposition in a coal gasification process

DOEpatents

Wilson, M.W.

1987-03-23

The gasification of coal in the presence of steam and oxygen is significantly enhanced by introducing a thermochemical water- splitting agent such as sulfuric acid, into the gasifier for decomposing the steam to provide additional oxygen and hydrogen usable in the gasification process for the combustion of the coal and enrichment of the gaseous gasification products. The addition of the water-splitting agent into the gasifier also allows for the operation of the reactor at a lower temperature.

Chemical Looping Autothermal Reforming at a 120 kW Pilot Rig

NASA Astrophysics Data System (ADS)

Bofhàr-Nordenkampf, Johannes; Pröll, Tobias; Kolbitsch, Philipp; Hofbauer, Hermann

Chemical looping with selective oxygen transport allows two step combustion or autothermal reforming without mixing of fuel and air. The reactor system consists of two reactors, an air reactor and a fuel reactor with a suitable oxygen carrier that transports the necessary oxygen for operation. In the present study, a highly active nickel based oxygen carrier is tested in a novel dual circulating fluidized bed (DCFB) system at a scale of 120 kW fuel power. The mean particle size of the oxygen carrier is 120 μm and the pilot rig is fueled with natural gas. For the investigated oxygen carrier high CH4 conversion is achieved. Air/fuel ratio is varied at three different fuel reactor temperatures. For chemical looping reforming one can observe synthesis gas composition close to thermodynamic equilibrium. In spite of the fact that no additional steam has been added to the fuel besides the one present through steam fluidization of the loop seals, coke formation does not occur at global stoichiometric air/fuel ratios above 0.46.

Properties of amylose complexes with hexadecyl amine and its hydrochloride salt prepared by steam jet cooking

USDA-ARS?s Scientific Manuscript database

Steam jet cooking of starch is an effective, commercially scalable method of preparing amylose for complexing with a variety of ligands. Previous work has shown that dispersions of amylose complexes prepared with fatty acids (such as palmitic) formed a variety of spherulites when cooled under diffe...

Downhole steam generator with improved preheating, combustion, and protection features

DOEpatents

Fox, R.L.

1981-01-07

For tertiary oil recovery, a downhole steam generator is designed which provides for efficient counterflow cooling of the combustion chamber walls and preheating of the fuel and water. Pressure-responsive doors are provided for closing and opening the outlet in response to flameout, thereby preventing flooding of the combustion chamber. (DLC)

Ultra-High Temperature Steam Corrosion of Complex Silicates for Nuclear Applications: A Computational Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rashkeev, Sergey N.; Glazoff, Michael V.; Tokuhiro, Akira

2014-01-01

Stability of materials under extreme conditions is an important issue for safety of nuclear reactors. Presently, silicon carbide (SiC) is being studied as a cladding material candidate for fuel rods in boiling-water and pressurized water-cooled reactors (BWRs and PWRs) that would substitute or modify traditional zircaloy materials. The rate of corrosion of the SiC ceramics in hot vapor environment (up to 2200 degrees C) simulating emergency conditions of light water reactor (LWR) depends on many environmental factors such as pressure, temperature, viscosity, and surface quality. Using the paralinear oxidation theory developed for ceramics in the combustion reactor environment, we estimatedmore » the corrosion rate of SiC ceramics under the conditions representing a significant power excursion in a LWR. It was established that a significant time – at least 100 h – is required for a typical SiC braiding to significantly degrade even in the most aggressive vapor environment (with temperatures up to 2200 °C) which is possible in a LWR at emergency condition. This provides evidence in favor of using the SiC coatings/braidings for additional protection of nuclear reactor rods against off-normal material degradation during power excursions or LOCA incidents. Additionally, we discuss possibilities of using other silica based ceramics in order to find materials with even higher corrosion resistance than SiC. In particular, we found that zircon (ZrSiO4) is also a very promising material for nuclear applications. Thermodynamic and first-principles atomic-scale calculations provide evidence of zircon thermodynamic stability in aggressive environments at least up to 1535 degrees C.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kreitman, Paul J.; Sirianni, Steve R.; Pillard, Mark M.

Entergy recently performed an Extended Power Up-rate (EPU) on their Grand Gulf Nuclear Station, near Port Gibson, Mississippi. To support the EPU, a new Steam Dryer Assembly was installed during the last refueling outage. Due to limited access into the containment, the large Replacement Steam Dryer (RSD) had to be brought into the containment in pieces and then final assembly was completed on the refueling floor before installation into the reactor. Likewise, the highly contaminated Original Steam Dryer (OSD) had to be segmented into manageable sections, loaded into specially designed shielded containers, and rigged out of containment where they willmore » be safely stored until final disposal is accomplished at an acceptable waste repository. Westinghouse Nuclear Services was contracted by Entergy to segment, package and remove the OSD from containment. This work was performed on critical path during the most recent refueling outage. The segmentation was performed underwater to minimize radiation exposure to the workers. Special hydraulic saws were developed for the cutting operations based on Westinghouse designs previously used in Sweden to segment ABB Reactor Internals. The mechanical cutting method was selected because of its proven reliability and the minimal cutting debris that is generated by the process. Maintaining stability of the large OSD sections during cutting was accomplished using a custom built support stand that was installed into the Moisture Separator Pool after the Moisture Separator was installed back in the reactor vessel. The OSD was then moved from the Steam Dryer Pool to the Moisture Separator Pool for segmentation. This scenario resolved the logistical challenge of having two steam dryers and a moisture separator in containment simultaneously. A water filtration/vacuum unit was supplied to maintain water clarity during the cutting and handling operations and to collect the cutting chips. (authors)« less

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 disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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 disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

Additive Manufacturing of Catalyst Substrates for Steam-Methane Reforming

NASA Astrophysics Data System (ADS)

Kramer, Michelle; McKelvie, Millie; Watson, Matthew

2018-01-01

Steam-methane reforming is a highly endothermic reaction, which is carried out at temperatures up to 1100 °C and pressures up to 3000 kPa, typically with a Ni-based catalyst distributed over a substrate of discrete alumina pellets or beads. Standard pellet geometries (spheres, hollow cylinders) limit the degree of mass transfer between gaseous reactants and catalyst. Further, heat is supplied to the exterior of the reactor wall, and heat transfer is limited due to the nature of point contacts between the reactor wall and the substrate pellets. This limits the degree to which the process can be intensified, as well as limiting the diameter of the reactor wall. Additive manufacturing now gives us the capability to design structures with tailored heat and mass transfer properties, not only within the packed bed of the reactor, but also at the interface between the reactor wall and the packed bed. In this work, the use of additive manufacturing to produce monolithic-structured catalyst substrate models, made from acrylonitrile-butadiene-styrene, with enhanced conductive heat transfer is described. By integrating the reactor wall into the catalyst substrate structure, the effective thermal conductivity increased by 34% from 0.122 to 0.164 W/(m K).

Feedwater temperature control methods and systems

DOEpatents

Moen, Stephan Craig; Noonan, Jack Patrick; Saha, Pradip

2014-04-22

A system for controlling the power level of a natural circulation boiling water nuclear reactor (NCBWR) is disclosed. The system, in accordance with an example embodiment of the present invention, may include a controller configured to control a power output level of the NCBWR by controlling a heating subsystem to adjust a temperature of feedwater flowing into an annulus of the NCBWR. The heating subsystem may include a steam diversion line configured to receive steam generated by a core of the NCBWR and a steam bypass valve configured to receive commands from the controller to control a flow of the steam in the steam diversion line, wherein the steam received by the steam diversion line has not passed through a turbine. Additional embodiments of the invention may include a feedwater bypass valve for controlling an amount of flow of the feedwater through a heater bypass line to the annulus.

Apparatus for solar coal gasification

DOEpatents

Gregg, D.W.

1980-08-04

Apparatus for using focused solar radiation to gasify coal and other carbonaceous materials is described. Incident solar radiation is focused from an array of heliostats through a window onto the surface of a moving bed of coal, contained within a gasification reactor. The reactor is designed to minimize contact between the window and solids in the reactor. Steam introduced into the gasification reactor reacts with the heated coal to produce gas consisting mainly of carbon monoxide and hydrogen, commonly called synthesis gas, which can be converted to methane, methanol, gasoline, and other useful products. One of the novel features of the invention is the generation of process steam in one embodiment at the rear surface of a secondary mirror used to redirect the focused sunlight. Another novel feature of the invention is the location and arrangement of the array of mirrors on an inclined surface (e.g., a hillside) to provide for direct optical communication of said mirrors and the carbonaceous feed without a secondary redirecting mirror.

DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

DOEpatents

Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

1962-08-14

A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

ASME Material Challenges for Advanced Reactor Concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piyush Sabharwall; Ali Siahpush

2013-07-01

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

Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Afifah, Maryam, E-mail: maryam.afifah210692@gmail.com; Su’ud, Zaki; Miura, Ryosuke

2015-09-30

Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design.more » The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.« less

NEUTRONIC REACTORS

DOEpatents

Vernon, H.C.

1959-01-13

A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

Emergency Cooling of Nuclear Power Plant Reactors With Heat Removal By a Forced-Draft Cooling Tower

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murav’ev, V. P., E-mail: murval1@mail.ru

The feasibility of heat removal during emergency cooling of a reactor by a forced-draft cooling tower with accumulation of the peak heat release in a volume of precooled water is evaluated. The advantages of a cooling tower over a spray cooling pond are demonstrated: it requires less space, consumes less material, employs shorter lines in the heat removal system, and provides considerably better protection of the environment from wetting by entrained moisture.

158. ARAIII Reactor building (ARA608) Secondary cooling loop and piping ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

158. ARA-III Reactor building (ARA-608) Secondary cooling loop and piping plan. This drawing was selected as a typical example of piping arrangements within reactor building. Aerojet/general 880-area/GCRE-608-P-16. Date: February 1958. INeel index code no. 063-0608-50-013-102641. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Method and apparatus for a catalytic firebox reactor

DOEpatents

Smith, Lance L.; Etemad, Shahrokh; Ulkarim, Hasan; Castaldi, Marco J.; Pfefferle, William C.

2001-01-01

A catalytic firebox reactor employing an exothermic catalytic reaction channel and multiple cooling conduits for creating a partially reacted fuel/oxidant mixture. An oxidation catalyst is deposited on the walls forming the boundary between the multiple cooling conduits and the exothermic catalytic reaction channel, on the side of the walls facing the exothermic catalytic reaction channel. This configuration allows the oxidation catalyst to be backside cooled by any fluid passing through the cooling conduits. The heat of reaction is added to both the fluid in the exothermic catalytic reaction channel and the fluid passing through the cooling conduits. After discharge of the fluids from the exothermic catalytic reaction channel, the fluids mix to create a single combined flow. A further innovation in the reactor incorporates geometric changes in the exothermic catalytic reaction channel to provide streamwise variation of the velocity of the fluids in the reactor.

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, M.M.; Lau, L.K.; Schulz, T.L.

1993-12-14

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

Method and apparatus for enhancing reactor air-cooling system performance

DOEpatents

Hunsbedt, A.

1996-03-12

An enhanced decay heat removal system is disclosed for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer. 6 figs.

Evaluation of the Performance of O-rings Made with Different Elastomeric Polymers in Simulated Geothermal Environments at 300°C

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sugama, Toshifumi; Pyatina, Tatiana; Redline, Erica Marie

2014-12-01

This paper aims to evaluate the survival of O-rings made with six different elastomeric polymers, EPDM, type I- and II-FKM, FEPM, FFKM, and FSR, in five different simulated geothermal environments at 300°C. It further defines the relative strengths and weaknesses of the materials in each environment. The environments tested were: 1) non-aerated steam-cooling cycles, 2) aerated steam-cooling cycles, 3) water-based drilling fluid, 4) CO2-rich geo-brine fluid, and, 5) heat-cool water quenching cycles. Following exposure, the extent of oxidation, oxidationinduced degradation, thermal behaviors, micro-defects, permeation depths of ionic species present in environments throughout the O-ring, silicate-related scale-deposition, and changes in mechanicalmore » properties were assessed.« less

PROCESS WATER BUILDING, TRA605. FLASH EVAPORATOR, CONDENSER (PROJECT FROM EVAPORATOR), ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PROCESS WATER BUILDING, TRA-605. FLASH EVAPORATOR, CONDENSER (PROJECT FROM EVAPORATOR), AND STEAM EJECTOR (ALONG REAR WALL). INL NEGATIVE NO. 4377. M.H. Bartz, Photographer, 3/5/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Method and apparatus for producing thermal vapor stream

DOEpatents

Cradeur, Robert R.; Sperry, John S.; Krajicek, Richard W.

1979-01-01

Method and apparatus for producing a thermal vapor stream for injecting into a subterranean formation for the recovery of liquefiable minerals therefrom, including a pressure vessel containing a high pressure combustion chamber for producing a heating gas for introduction into a heating gas injector. The heating gas injector is partly immersed in a steam generating section of the pressure vessel such that the heating gas is passed through the steam generating section to produce steam and combustion products which are directed between the pressure vessel and the combustion chamber for simultaneously cooling of the combustion chamber by further heating of the steam and combustion gases.

Method and apparatus for obtaining enhanced production rate of thermal chemical reactions

DOEpatents

Tonkovich, Anna Lee Y.; Wang, Yong; Wegeng, Robert S.; Gao, Yufei

2003-09-09

Reactors and processes are disclosed that can utilize high heat fluxes to obtain fast, steady-state reaction rates. Porous catalysts used in conjunction with microchannel reactors to obtain high rates of heat transfer are also disclosed. Reactors and processes that utilize short contact times, high heat flux and low pressure drop are described. Improved methods of steam reforming are also provided.

Method and apparatus for obtaining enhanced production rate of thermal chemical reactions

DOEpatents

Tonkovich, Anna Lee Y [Pasco, WA; Wang, Yong [Richland, WA; Wegeng, Robert S [Richland, WA; Gao, Yufei [Kennewick, WA

2006-05-16

Reactors and processes are disclosed that can utilize high heat fluxes to obtain fast, steady-state reaction rates. Porous catalysts used in conjunction with microchannel reactors to obtain high rates of heat transfer are also disclosed. Reactors and processes that utilize short contact times, high heat flux and low pressure drop are described. Improved methods of steam reforming are also provided.

Thermal gain of CHP steam generator plants and heat supply systems

NASA Astrophysics Data System (ADS)

Ziganshina, S. K.; Kudinov, A. A.

2016-08-01

Heating calculation of the surface condensate heat recovery unit (HRU) installed behind the BKZ-420-140 NGM boiler resulting in determination of HRU heat output according to fire gas value parameters at the heat recovery unit inlet and its outlet, heated water quantity, combustion efficiency per boiler as a result of installation of HRU, and steam condensate discharge from combustion products at its cooling below condensing point and HRU heat exchange area has been performed. Inspection results of Samara CHP BKZ-420-140 NGM power boilers and field tests of the surface condensate heat recovery unit (HRU) made on the bimetal calorifier base KCk-4-11 (KSk-4-11) installed behind station no. 2 Ulyanovsk CHP-3 DE-10-14 GM boiler were the basis of calculation. Integration of the surface condensation heat recovery unit behind a steam boiler rendered it possible to increase combustion efficiency and simultaneously decrease nitrogen oxide content in exit gases. Influence of the blowing air moisture content, the excess-air coefficient in exit gases, and exit gases temperature at the HRU outlet on steam condensate amount discharge from combustion products at its cooling below condensing point has been analyzed. The steam condensate from HRU gases is offered as heat system make-up water after degasification. The cost-effectiveness analysis of HRU installation behind the Samara CHP BKZ-420-140 NGM steam boiler with consideration of heat energy and chemically purified water economy has been performed. Calculation data for boilers with different heat output has been generalized.

Benchmark Simulation of Natural Circulation Cooling System with Salt Working Fluid Using SAM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahmed, K. K.; Scarlat, R. O.; Hu, R.

Liquid salt-cooled reactors, such as the Fluoride Salt-Cooled High-Temperature Reactor (FHR), offer passive decay heat removal through natural circulation using Direct Reactor Auxiliary Cooling System (DRACS) loops. The behavior of such systems should be well-understood through performance analysis. The advanced system thermal-hydraulics tool System Analysis Module (SAM) from Argonne National Laboratory has been selected for this purpose. The work presented here is part of a larger study in which SAM modeling capabilities are being enhanced for the system analyses of FHR or Molten Salt Reactors (MSR). Liquid salt thermophysical properties have been implemented in SAM, as well as properties ofmore » Dowtherm A, which is used as a simulant fluid for scaled experiments, for future code validation studies. Additional physics modules to represent phenomena specific to salt-cooled reactors, such as freezing of coolant, are being implemented in SAM. This study presents a useful first benchmark for the applicability of SAM to liquid salt-cooled reactors: it provides steady-state and transient comparisons for a salt reactor system. A RELAP5-3D model of the Mark-1 Pebble-Bed FHR (Mk1 PB-FHR), and in particular its DRACS loop for emergency heat removal, provides steady state and transient results for flow rates and temperatures in the system that are used here for code-to-code comparison with SAM. The transient studied is a loss of forced circulation with SCRAM event. To the knowledge of the authors, this is the first application of SAM to FHR or any other molten salt reactors. While building these models in SAM, any gaps in the code’s capability to simulate such systems are identified and addressed immediately, or listed as future improvements to the code.« less

COOLING TOWER PUMP HOUSE, TRA606. THREE OF SIX SECTIONS OF ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

COOLING TOWER PUMP HOUSE, TRA-606. THREE OF SIX SECTIONS OF COOLING TOWER ARE VISIBLE ABOVE RAILING. PUMP HOUSE IN FOREGROUND IS ON SOUTH SIDE OF COOLING TOWER. NOTE THREE PIPES TAKING WATER FROM PUMP HOUSE TO HOT DECK OF COOLING TOWER. EMERGENCY WATER SUPPLY TOWER IS ALSO IN VIEW. INL NEGATIVE NO. 6197. Unknown Photographer, 6/27/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Saturation meter

DOEpatents

Gregurech, S.

1984-08-01

A saturation meter for use in a pressurized water reactor plant comprising a differential pressure transducer having a first and second pressure sensing means and an alarm. The alarm is connected to the transducer and is preset to activate at a level of saturation prior to the formation of a steam void in the reactor vessel.

Sources and potential application of waste heat utilization at a gas processing facility

NASA Astrophysics Data System (ADS)

Alshehhi, Alyas Ali

Waste heat recovery (WHR) has the potential to significantly improve the efficiency of oil and gas plants, chemical and other processing facilities, and reduce their environmental impact. In this Thesis a comprehensive energy audit at Abu Dhabi Gas Industries Ltd. (GASCO) ASAB gas processing facilities is undertaken to identify sources of waste heat and evaluate their potential for on-site recovery. Two plants are considered, namely ASAB0 and ASAB1. Waste heat evaluation criteria include waste heat grade (i.e., temperature), rate, accessibility (i.e., proximity) to potential on-site waste heat recovery applications, and potential impact of recovery on installation performance and safety. The operating parameters of key waste heat source producing equipment are compiled, as well as characteristics of the waste heat streams. In addition, potential waste heat recovery applications and strategies are proposed, focusing on utilities, i.e., enhancement of process cooling/heating, electrical/mechanical power generation, and steam production. The sources of waste heat identified at ASAB facilities consist of gas turbine and gas generator exhaust gases, flared gases, excess propane cooling capacity, excess process steam, process gas air-cooler heat dissipation, furnace exhaust gases and steam turbine outlet steam. Of the above waste heat sources, exhaust gases from five gas turbines and one gas generator at ASAB0 plant, as well as from four gas turbines at ASAB1 plant, were found to meet the rate (i.e., > 1 MW), grade (i.e., > 180°C), accessibility (i.e., < 50 m from potential on-site WHR applications) and minimal impact criteria on the performance and safety of existing installations, for potential waste heat recovery. The total amount of waste heat meeting these criteria were estimated at 256 MW and 289 MW at ASAB0 and ASAB1 plants, respectively, both of which are substantial. Of the 289 MW waste generated at ASAB1, approximately 173 MW are recovered by waste heat recovery steam generators (WHRSGs), leaving 116 MW unutilized. The following strategies were developed to recover the above waste heat. At ASAB0, it is proposed that exhaust gases from all five gas turbines be used to power a WHRSG. The steam generated by the WHRSG would both i) drive an absorption refrigeration unit for gas turbine inlet air cooling, which would result in additional electric or mechanical power generation, and pre-cooling of process gas, which could reduce the need for or eliminate air coolers, as well as reduce propane chiller load, and ii) serve for heating of lean gas, which would reduce furnace load. At ASAB1, it is proposed that exhaust gases from all four gas turbines be used to generate steam in WHRSG that would drive an absorption refrigeration unit for either gas turbine inlet air cooling for additional electric or mechanical power generation, or pre-cooling of process gas to eliminate air-coolers and reduce propane chiller cooling load. Considering the smaller amount of waste heat available at ASAB1 (116 MW) relative to ASAB0 (237 MW), these above two recovery options could not be implemented simultaneously at ASAB0. To permit the detailed design and techno-economic feasibility evaluation of the proposed waste heat recovery strategies in a subsequent study, the cooling loads and associated electric power consumption of ASAB0 process gas air-coolers were estimated at 21 MW and 1.9 MW, respectively, and 67 MW and 2.2 MW, respectively for ASAB1 plant. In addition, the heating loads and fuel consumption of ASAB0 furnaces used for lean gas re-generation were estimated at 24 MW and 0.0653 MMSCMD, respectively. In modeling work undertaken in parallel with this study at the Petroleum Institute, the waste heat recovery strategies proposed here were found to be thermodynamically and economically feasible, and to lead to substantial energy and cost savings, hence environmental benefits.

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

2013-09-25

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in amore » remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.« less

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

ScienceCinema

None

2018-01-16

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

Calculation of the Oxide Growth Rate of 2-1/4 Cr - 1 Mo Steel in Air and the Subsequent Measurement of the Strain Required to Crack the Oxide.

DTIC Science & Technology

1980-09-01

freezing points were measured using the prepared thermocouple and reader. It was found that over this temperature range, temperatures were measured...Manning and E. Metcalfe, "Oxidation of Ferritic Steels in Steam," International Conference on Ferritic Steels for Fast Reactor Steam Generators

Influence of geometry on pressure and velocity distribution in packed-bed methanol steam reforming reactor

NASA Astrophysics Data System (ADS)

Ivanović, Ivana; Sedmak, Aleksandar; Milošević, Miloš; Cvetković, Ivana; Pohar, Andrej; Likozar, Blaž

2017-07-01

The main tasks of this research is to propose several changes in the packed bed micro methanol steam reformer geometry in order to ensure its performance. The reformer is an integral part of the existing indirect internal reforming high temperature PEMFC and most of its geometry is already defined. The space for remodeling is very limited.

Modeling and Comparison of Options for the Disposal of Excess Weapons Plutonium in Russia

DTIC Science & Technology

2002-04-01

fuel LWR cooling time LWR Pu load rate LWR net destruction frac ~ LWR reactors op life mox core frac Excess Separated Pu HTGR Cycle Pu in Waste LWR MOX...reflecting the cycle used in this type of reactor. For the HTGR , the entire core consists of plutonium fuel , therefore a core fraction is not specified...cooling time Time spent fuel unloaded from HTGR reactor must cool before permanently stored 3 years Mox core fraction Fraction of

Impact of irrigation flow rate and intrapericardial fluid on cooled-tip epicardial radiofrequency ablation.

PubMed

Aryana, Arash; O'Neill, Padraig Gearoid; Pujara, Deep K; Singh, Steve K; Bowers, Mark R; Allen, Shelley L; d'Avila, André

2016-08-01

The optimal irrigation flow rate (IFR) during epicardial radiofrequency (RF) ablation has not been established. This study specifically examined the impact of IFR and intrapericardial fluid (IPF) accumulation during epicardial RF ablation. Altogether, 452 ex vivo RF applications (10 g for 60 seconds) delivered to the epicardial surface of bovine myocardium using 3 open-irrigated ablation catheters (ThermoCool SmartTouch, ThermoCool SmartTouch-SF, and FlexAbility) and 50 in vivo RF applications delivered (ThermoCool SmartTouch-SF) in 4 healthy adult swine in the presence or absence of IPF were examined. Ex vivo, RF was delivered at low (≤3 mL/min), reduced (5-7 mL/min), and high (≥10 mL/min) IFRs using intermediate (25-35 W) and high (35-45 W) power. In vivo, applications were delivered (at 9.3 ± 2.2 g for 60 seconds at 39 W) using reduced (5 mL/min) and high (15 mL/min) IFRs. Ex vivo, surface lesion diameter inversely correlated with IFR, whereas maximum lesion diameter and depth did not differ. While steam pops occurred more frequently at low IFR using high power (ThermoCool SmartTouch and ThermoCool SmartTouch-SF), tissue disruption was rare and did not vary with IFR. In vivo, charring/steam pop was not detected. Although there were no discernible differences in lesion size with IFR, surface lesion diameter, maximum diameter, depth, and volume were all smaller in the presence of IPF at both IFRs. Cooled-tip epicardial RF ablation created using reduced IFRs (5-7 mL/min) yields lesion sizes similar to those created using high IFRs (≥10 mL/min) without an increase in steam pop/tissue disruption, whereas the presence of IPF significantly reduces the lesion size. Copyright © 2016 Heart Rhythm Society. Published by Elsevier Inc. All rights reserved.

Heat exchanger with auxiliary cooling system

DOEpatents

Coleman, John H.

1980-01-01

A heat exchanger with an auxiliary cooling system capable of cooling a nuclear reactor should the normal cooling mechanism become inoperable. A cooling coil is disposed around vertical heat transfer tubes that carry secondary coolant therethrough and is located in a downward flow of primary coolant that passes in heat transfer relationship with both the cooling coil and the vertical heat transfer tubes. A third coolant is pumped through the cooling coil which absorbs heat from the primary coolant which increases the downward flow of the primary coolant thereby increasing the natural circulation of the primary coolant through the nuclear reactor.

Material Issues of Blanket Systems for Fusion Reactors - Compatibility with Cooling Water -

NASA Astrophysics Data System (ADS)

Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro

Environmental assisted cracking (EAC) is one of the material issues for the reactor core components of light water power reactors(LWRs). Much experience and knowledge have been obtained about the EAC in the LWR field. They will be useful to prevent the EAC of water-cooled blanket systems of fusion reactors. For the austenitic stainless steels and the reduced-activation ferritic/martensitic steels, they clarifies that the EAC in a water-cooled blanket does not seem to be acritical issue. However, some uncertainties about influences on water temperatures, water chemistries and stress conditions may affect on the EAC. Considerations and further investigations elucidating the uncertainties are discussed.

Cesium vapor cycle for an advanced LMFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fraas, A.P.

1975-01-01

A review indicates that a cesium vapor topping cycle appears attractive for use in the intermediate fluid circuit of an advanced LMFBR designed for a reactor outlet temperature of 1250$sup 0$F or more and would have the following advantages: (1) it would increase the thermal efficiency by about 5 to 10 points (from approximately 40 percent to approximately 45 to 50 percent) thus reducing the amount of waste heat rejected to the environment by 15 to 30 percent. (2) the higher thermal efficiency should reduce the overall capital cost of the reactor plant in dollars per kilowatt. (3) the cesiummore » can be distilled out of the intermediate fluid circuit to leave it bone-dry, thus greatly reducing the time and cost of maintenance work (particularly for the steam generator). (4) the large volume and low pressure of the cesium vapor region in the cesium condenser-steam generator greatly reduces the magnitude of pressure fluctuations that might occur in the event of a leak in a steam generator tube, and the characteristics inherent in a condenser make it easy to design for rapid concentration of any noncondensibles that may form as a consequence of a steam leak into the cesium region so that a steam leak can be detected easily in the very early stages of its development. (auth)« less

78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0237] Cost-Benefit Analysis for Radwaste Systems for Light... (RG) 1.110, ``Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors... components for light water nuclear power reactors. ADDRESSES: Please refer to Docket ID NRC-2013-0237 when...

Application of a Self-Actuating Shutdown System (SASS) to a Gas-Cooled Fast Reactor (GCFR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Germer, J.H.; Peterson, L.F.; Kluck, A.L.

1980-09-01

The application of a SASS (Self-Actuated Shutdown System) to a GCFR (Gas-Cooled Fast Reactor) is compared with similar systems designed for an LMFBR (Liquid Metal Fast Breeder Reactor). A comparison of three basic SASS concepts is given: hydrostatic holdup, fluidic control, and magnetic holdup.

PROCESS FOR COOLING A NUCLEAR REACTOR

DOEpatents

Borst, L.B.

1962-12-11

This patent relates to the operation of a reactor cooled by liquid sulfur dioxide. According to the invention the pressure on the sulfur dioxide in the reactor is maintained at least at the critical pressure of the sulfur dioxide. Heating the sulfur dioxide to its critical temperature results in vaporization of the sulfur dioxide without boiling. (AEC)

78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

...The U.S. Nuclear Regulatory Commission (NRC) is issuing a revision to regulatory guide (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors.'' This RG is being revised to incorporate guidance for preoperational testing of new pressurized water reactor (PWR) designs.

Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

2017-12-01

The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Fielding, R.; Gan, J.

2007-09-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

LIQUID METAL REACTOR COOLING SYSTEMS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aberdam, M.; Gros, G.

1965-02-01

This report is part of a series of bibliographies. The specific purpose of this report is to describe the various elements of the cooling systems in the principal liquid-metal-cooled reactors now operating, being contsructed, or in the design stage. The information given is drawn from reports or publicatios received during or before September 1964.

Bi-reforming of methane from any source with steam and carbon dioxide exclusively to metgas (CO-2H2) for methanol and hydrocarbon synthesis.

PubMed

Olah, George A; Goeppert, Alain; Czaun, Miklos; Prakash, G K Surya

2013-01-16

A catalyst based on nickel oxide on magnesium oxide (NiO/MgO) thermally activated under hydrogen is effective for the bi-reforming with steam and CO(2) (combined steam and dry reforming) of methane as well as natural gas in a tubular flow reactor at elevated pressures (5-30 atm) and temperatures (800-950 °C). By adjusting the CO(2)-to-steam ratio in the gas feed, the H(2)/CO ratio in the produced syn-gas could be easily adjusted in a single step to the desired value of 2 for methanol and hydrocarbon synthesis.

NUCLEAR FLASH TYPE STEAM GENERATOR

DOEpatents

Johns, F.L.; Gronemeyer, E.C.; Dusbabek, M.R.

1962-09-01

A nuclear steam generating apparatus is designed so that steam may be generated from water heated directly by the nuclear heat source. The apparatus comprises a pair of pressure vessels mounted one within the other, the inner vessel containing a nuclear reactor heat source in the lower portion thereof to which water is pumped. A series of small ports are disposed in the upper portion of the inner vessel for jetting heated water under pressure outwardly into the atmosphere within the interior of the outer vessel, at which time part of the jetted water flashes into steam. The invention eliminates the necessity of any intermediate heat transfer medium and components ordinarily required for handling that medium. (AEC)

LCRE and SNAP 50-DR-1 programs. Engineering and progress report, April 1, 1963--June 30, 1963

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

BS>Declassified 6 Sep 1973. Information is presented concerning the LCRE kinetics, auxiliary systems, fuel, primary cooling system components, instrumentation, secondary cooling system, materials development, and fabrication; and SNAP-50/SPUR kinetics, fuel, primary system pump, steam generator, and materials development. (DCC)

Estimation of transient heat flux density during the heat supply of a catalytic wall steam methane reformer

NASA Astrophysics Data System (ADS)

Settar, Abdelhakim; Abboudi, Saïd; Madani, Brahim; Nebbali, Rachid

2018-02-01

Due to the endothermic nature of the steam methane reforming reaction, the process is often limited by the heat transfer behavior in the reactors. Poor thermal behavior sometimes leads to slow reaction kinetics, which is characterized by the presence of cold spots in the catalytic zones. Within this framework, the present work consists on a numerical investigation, in conjunction with an experimental one, on the one-dimensional heat transfer phenomenon during the heat supply of a catalytic-wall reactor, which is designed for hydrogen production. The studied reactor is inserted in an electric furnace where the heat requirement of the endothermic reaction is supplied by electric heating system. During the heat supply, an unknown heat flux density, received by the reactive flow, is estimated using inverse methods. In the basis of the catalytic-wall reactor model, an experimental setup is engineered in situ to measure the temperature distribution. Then after, the measurements are injected in the numerical heat flux estimation procedure, which is based on the Function Specification Method (FSM). The measured and estimated temperatures are confronted and the heat flux density which crosses the reactor wall is determined.

Fuel processing in integrated micro-structured heat-exchanger reactors

NASA Astrophysics Data System (ADS)

Kolb, G.; Schürer, J.; Tiemann, D.; Wichert, M.; Zapf, R.; Hessel, V.; Löwe, H.

Micro-structured fuel processors are under development at IMM for different fuels such as methanol, ethanol, propane/butane (LPG), gasoline and diesel. The target application are mobile, portable and small scale stationary auxiliary power units (APU) based upon fuel cell technology. The key feature of the systems is an integrated plate heat-exchanger technology which allows for the thermal integration of several functions in a single device. Steam reforming may be coupled with catalytic combustion in separate flow paths of a heat-exchanger. Reactors and complete fuel processors are tested up to the size range of 5 kW power output of a corresponding fuel cell. On top of reactor and system prototyping and testing, catalyst coatings are under development at IMM for numerous reactions such as steam reforming of LPG, ethanol and methanol, catalytic combustion of LPG and methanol, and for CO clean-up reactions, namely water-gas shift, methanation and the preferential oxidation of carbon monoxide. These catalysts are investigated in specially developed testing reactors. In selected cases 1000 h stability testing is performed on catalyst coatings at weight hourly space velocities, which are sufficiently high to meet the demands of future fuel processing reactors.

Vacuum chamber with a supersonic flow aerodynamic window

DOEpatents

Hanson, Clark L.

1982-01-01

A supersonic flow aerodynamic window, whereby a steam ejector situated in a primary chamber at vacuum exhausts superheated steam toward an orifice to a region of higher pressure, creating a barrier to the gas in the region of higher pressure which attempts to enter through the orifice. In a mixing chamber outside and in fluid communication with the primary chamber, superheated steam and gas are combined into a mixture which then enters the primary chamber through the orifice. At the point of impact of the ejector/superheated steam and the incoming gas/superheated steam mixture, a barrier is created to the gas attempting to enter the ejector chamber. This barrier, coupled with suitable vacuum pumping means and cooling means, serves to keep the steam ejector and primary chamber at a negative pressure, even though the primary chamber has an orifice to a region of higher pressure.

Vacuum chamber with a supersonic-flow aerodynamic window

DOEpatents

Hanson, C.L.

1980-10-14

A supersonic flow aerodynamic window is disclosed whereby a steam ejector situated in a primary chamber at vacuum exhausts superheated steam toward an orifice to a region of higher pressure, creating a barrier to the gas in the region of higher pressure which attempts to enter through the orifice. In a mixing chamber outside and in fluid communication with the primary chamber, superheated steam and gas are combined into a mixture which then enters the primary chamber through the orifice. At the point of impact of the ejector/superheated steam and the incoming gas/superheated steam mixture, a barrier is created to the gas attempting to enter the ejector chamber. This barrier, coupled with suitable vacuum pumping means and cooling means, serves to keep the steam ejector and primary chamber at a negative pressure, even though the primary chamber has an orifice to a region of higher pressure.

Biophysics and clinical utility of irrigated-tip radiofrequency catheter ablation.

PubMed

Houmsse, Mahmoud; Daoud, Emile G

2012-01-01

Catheter ablation by radiofrequency (RF) energy has successfully eliminated cardiac tachyarrhythmias. RF ablation lesions are created by thermal energy. Electrode catheters with 4-mm-tips have been adequate to ablate arrhythmias located near the endocardium; however, the 4-mm-tip electrode does not readily ablate deeper tachyarrhythmia substrate. With 8- and 10-mm-tip RF electrodes, ablation lesions were larger; yet, these catheters are associated with increased risk for coagulum, char and thrombus formation, as well as myocardial steam rupture. Cooled-tip catheter technology was designed to cool the electrode tip, prevent excessive temperatures at the electrode tip-tissue interface, and thus allow continued delivery of RF current into the surrounding tissue. This ablation system creates larger and deeper ablation lesions and minimizes steam pops and thrombus formation. The purpose of this article is to review cooled-tip RF ablation biophysics and outcomes of clinical studies as well as to discuss future technological improvements.

Passive heat transfer means for nuclear reactors

DOEpatents

Burelbach, James P.

1984-01-01

An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. Means such as shrouding normally isolated the secondary condensing section from effective heat transfer with the heat sink, but a sensor responds to overheat conditions of the reactor to open the shrouding, which thereby increases the cooling capacity of the heat pipe. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

Pilot-scale study on the acid-catalyzed steam explosion of rice straw using a continuous pretreatment system.

PubMed

Chen, Wen-Hua; Tsai, Chia-Chin; Lin, Chih-Feng; Tsai, Pei-Yuan; Hwang, Wen-Song

2013-01-01

A continuous acid-catalyzed steam explosion pretreatment process and system to produce cellulosic ethanol was developed at the pilot-scale. The effects of the following parameters on the pretreatment efficiency of rice straw feedstocks were investigated: the acid concentration, the reaction temperature, the residence time, the feedstock size, the explosion pressure and the screw speed. The optimal presteaming horizontal reactor conditions for the pretreatment process are as follows: 1.7 rpm and 100-110 °C with an acid concentration of 1.3% (w/w). An acid-catalyzed steam explosion is then performed in the vertical reactor at 185 °C for 2 min. Approximately 73% of the total saccharification yield was obtained after the rice straw was pretreated under optimal conditions and subsequent enzymatic hydrolysis at a combined severity factor of 0.4-0.7. Moreover, good long-term stability and durability of the pretreatment system under continuous operation was observed. Copyright © 2012 Elsevier Ltd. All rights reserved.

Enhanced simultaneous saccharification and fermentation of pretreated beech wood by in situ treatment with the white rot fungus Irpex lacteus in a membrane aerated biofilm reactor.

PubMed

Brethauer, Simone; Robert Lawrence, Shahab; Michael Hans-Peter, Studer

2017-08-01

The aim of the present study was to investigate the combination of steam pretreatment and biological treatment with lignin degrading fungal strains in order to enable efficient bioprocessing of beech wood to ethanol. In a sequential process of steam and fungal pretreatment followed by enzymatic hydrolysis, Irpex lacteus almost doubled the glucose yield for mildly pretreated beech wood, but could not improve yields for more severely pretreated substrates. However, when simultaneous saccharification and fermentation is combined with in situ I. lacteus treatment, which is enabled by the application of a membrane aerated biofilm reactor, ethanol yields of optimally steam pretreated beech could be improved from 65 to 80%. Generally, in situ fungal treatment during bioprocessing of lignocellulose is an interesting method to harness the versatile abilities of white rot fungi. Copyright © 2017 The Author(s). Published by Elsevier Ltd.. All rights reserved.

An Overview of INEL Fusion Safety R&D Facilities

NASA Astrophysics Data System (ADS)

McCarthy, K. A.; Smolik, G. R.; Anderl, R. A.; Carmack, W. J.; Longhurst, G. R.

1997-06-01

The Fusion Safety Program at the Idaho National Engineering Laboratory has the lead for fusion safety work in the United States. Over the years, we have developed several experimental facilities to provide data for fusion reactor safety analyses. We now have four major experimental facilities that provide data for use in safety assessments. The Steam-Reactivity Measurement System measures hydrogen generation rates and tritium mobilization rates in high-temperature (up to 1200°C) fusion relevant materials exposed to steam. The Volatilization of Activation Product Oxides Reactor Facility provides information on mobilization and transport and chemical reactivity of fusion relevant materials at high temperature (up to 1200°C) in an oxidizing environment (air or steam). The Fusion Aerosol Source Test Facility is a scaled-up version of VAPOR. The ion-implanta-tion/thermal-desorption system is dedicated to research into processes and phenomena associated with the interaction of hydrogen isotopes with fusion materials. In this paper we describe the capabilities of these facilities.

The influence of EI-21 redox ion-exchange resins on the secondary-coolant circuit water chemistry of vehicular nuclear power installations

NASA Astrophysics Data System (ADS)

Moskvin, L. N.; Rakov, V. T.

2015-06-01

The results obtained from testing the secondary-coolant circuit water chemistry of full-scale land-based prototype bench models of vehicular nuclear power installations equipped with water-cooled water-moderated and liquid-metal reactor plants are presented. The influence of copper-containing redox ionexchange resins intended for chemically deoxygenating steam condensate on the working fluid circulation loop's water chemistry is determined. The influence of redox ion-exchange resins on the water chemistry is evaluated by generalizing an array of data obtained in the course of extended monitoring using the methods relating to physicochemical analysis of the quality of condensate-feedwater path media and the methods relating to metallographic analysis of the state of a faulty steam generator's tube system surfaces. The deoxygenating effectiveness of the normal state turbine condensate vacuum deaeration system is experimentally determined. The refusal from applying redox ion-exchange resins in the condensate polishing ion-exchange filters is formulated based on the obtained data on the adverse effect of copper-containing redox ionexchange resins on the condensate-feedwater path water chemistry and based on the data testifying a sufficient effect from using the normal state turbine condensate vacuum deaeration system. Data on long-term operation of the prototype bench model of a vehicular nuclear power installation without subjecting the turbine condensate to chemical deoxygenation are presented.

IEA/SPS 500 kW distributed collector system

NASA Technical Reports Server (NTRS)

Neumann, T. W.; Hartman, C. D.

1980-01-01

Engineering studies for an International Energy Agency project for the design and construction of a 500 kW solar thermal electric power generation system of the distributed collector system (DCS) type are reviewed. The DCS system design consists of a mixed field of parabolic trough type solar collectors which are used to heat a thermal heat transfer oil. Heated oil is delivered to a thermocline storage tank from which heat is extracted and delivered to a boiler by a second heat transfer loop using the same heat transfer oil. Steam is generated in the boiler, expanded through a steam turbine, and recirculated through a condenser system cooled by a wet cooling tower.

Multitechnique characterisation of 304L surface states oxidised at high temperature in steam and air atmospheres

NASA Astrophysics Data System (ADS)

Mamede, Anne-Sophie; Nuns, Nicolas; Cristol, Anne-Lise; Cantrel, Laurent; Souvi, Sidi; Cristol, Sylvain; Paul, Jean-François

2016-04-01

In case of a severe accident occurring in a nuclear reactor, surfaces of the reactor coolant system (RCS), made of stainless steel (304L) rich in Cr (>10%) and Ni (8-12%), are oxidised. Fission products (FPs) are released from melt fuel and flow through the RCS. A part of them is deposited onto surfaces either by vapour condensation or by aerosol deposition mechanisms. To be able to understand the nature of interactions between these FPs and the RCS surfaces, a preliminary step is to characterize the RSC surface states in steam and air atmosphere at high temperatures. Pieces of 304L stainless steel have been treated in a flow reactor at two different temperatures (750 °C and 950 °C) for two different exposition times (24 h and 72 h). After surfaces analysing by a unique combination of surface analysis techniques (XPS, ToF-SIMS and LEIS), for 304L, the results show a deep oxide scale with multi layers and the outer layer is composed of chromium and manganese oxides. Oxide profiles differ in air or steam atmosphere. Fe2O3 oxide is observed but in minor proportion and in all cases no nickel is detected near the surface. Results obtained are discussed and compared with the literature data.

Parametric performance analysis of steam-injected gas turbine with a thermionic-energy-converter-lined combustor

NASA Technical Reports Server (NTRS)

Choo, Y. K.; Burns, R. K.

1982-01-01

The performance of steam-injected gas turbines having combustors lined with thermionic energy converters (STIG/TEC systems) was analyzed and compared with that of two baseline systems; a steam-injected gas turbine (without a TEC-lined combustor) and a conventional combined gas turbine/steam turbine cycle. Common gas turbine parameters were assumed for all of the systems. Two configurations of the STIG/TEC system were investigated. In both cases, steam produced in an exhaust-heat-recovery boiler cools the TEC collectors. It is then injected into the gas combustion stream and expanded through the gas turbine. The STIG/TEC system combines the advantage of gas turbine steam injection with the conversion of high-temperature combustion heat by TEC's. The addition of TEC's to the baseline steam-injected gas turbine improves both its efficiency and specific power. Depending on system configuration and design parameters, the STIG/TEC system can also achieve higher efficiency and specific power than the baseline combined cycle.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Merzari, E.; Yuan, Haomin; Kraus, A.

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable componentsmore » in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the design stage. Over the past five years, the Reactor Product Line has developed the integrated multi-physics code suite SHARP. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside the full reactor core or portions of it with a user-specified fidelity. In particular SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. These tools form a strong basis on which to build a flow-induced vibration modeling capability. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo aimed at simulating available FIV experiments in helical steam generators in the turbulent buffeting regime. In this regime one-way coupling is judged sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators at the low flows expected in integral PWRs. The legacy data is obtained from two datasets developed at Argonne and B&W.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

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.

Control rod system useable for fuel handling in a gas-cooled nuclear reactor

DOEpatents

Spurrier, Francis R.

1976-11-30

A control rod and its associated drive are used to elevate a complete stack of fuel blocks to a position above the core of a gas-cooled nuclear reactor. A fuel-handling machine grasps the control rod and the drive is unlatched from the rod. The stack and rod are transferred out of the reactor, or to a new location in the reactor, by the fuel-handling machine.

Transient Response to Rapid Cooling of a Stainless Steel Sodium Heat Pipe

NASA Technical Reports Server (NTRS)

Mireles, Omar R.; Houts, Michael G.

2011-01-01

Compact fission power systems are under consideration for use in long duration space exploration missions. Power demands on the order of 500 W, to 5 kW, will be required for up to 15 years of continuous service. One such small reactor design consists of a fast spectrum reactor cooled with an array of in-core alkali metal heat pipes coupled to thermoelectric or Stirling power conversion systems. Heat pipes advantageous attributes include a simplistic design, lack of moving parts, and well understood behavior. Concerns over reactor transients induced by heat pipe instability as a function of extreme thermal transients require experimental investigations. One particular concern is rapid cooling of the heat pipe condenser that would propagate to cool the evaporator. Rapid cooling of the reactor core beyond acceptable design limits could possibly induce unintended reactor control issues. This paper discusses a series of experimental demonstrations where a heat pipe operating at near prototypic conditions experienced rapid cooling of the condenser. The condenser section of a stainless steel sodium heat pipe was enclosed within a heat exchanger. The heat pipe - heat exchanger assembly was housed within a vacuum chamber held at a pressure of 50 Torr of helium. The heat pipe was brought to steady state operating conditions using graphite resistance heaters then cooled by a high flow of gaseous nitrogen through the heat exchanger. Subsequent thermal transient behavior was characterized by performing an energy balance using temperature, pressure and flow rate data obtained throughout the tests. Results indicate the degree of temperature change that results from a rapid cooling scenario will not significantly influence thermal stability of an operating heat pipe, even under extreme condenser cooling conditions.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gougar, Hans David

2015-10-01

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

Passive cooling system for nuclear reactor containment structure

DOEpatents

Gou, Perng-Fei; Wade, Gentry E.

1989-01-01

A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

Posttest data analysis and assessment of TRAC-BD1/MOD1 with data from a Full Integral Simulation Test (FIST) power transient experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wheatley, P.D.; Wagner, K.C.

The FIST power transient test 6PMC2 was analyzed to further the understanding of the FIST facility and provide an assessment of TRAC-BD1/MOD1. FIST power transient 6PMC2 investigated the thermal-hydraulic response following inadvertent closure of the main steam isolation valve and the subsequent failure of the reactor to scram. Failure of the high pressure core spray system was also assumed, resulting on only the reactor core isolation cooling flow for inventory makeup during the transient. The experiment was a sensitivity study with relatively high core power and low makeup rates. This study provides one of the first opportunities to assess TRAC-BD1/MOD1more » under power transient and natural circulation conditions with data from a facility with prototypical BWR geometry. The power transient test was analyzed with emphasis on the following phenomena; (a) the system pressure response, (b) the natural circulation flows and rates, and (c) the heater rod cladding temperature response. Based on the results of this study, TRAC-BD1/MOD1 can be expected to calculate the thermal-hydraulic behavior of a BWR during a power transient.« less

Capabilities Development for Transient Testing of Advanced Nuclear Fuels at TREAT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Woolstenhulme, N. E.; Baker, C. C.; Bess, J. D.

2016-09-01

The TREAT facility is a unique capability at the Idaho National Laboratory currently being prepared for resumption of nuclear transient testing. Accordingly, designs for several transient irradiation tests are being pursued to enable development of advanced nuclear fuels and materials. In addition to the reactor itself, the foundation for TREAT’s capabilities also include a suite of irradiation vehicles and supporting infrastructure to provide the desired specimen boundary conditions while supporting a variety of instrumentation needs. The challenge of creating these vehicles, especially since many of the modern data needs were not historically addressed in TREAT experiment vehicles, has necessitated amore » sizeable engineering effort. This effort is currently underway and maturing rapidly. This paper summarizes the status, future plans, and rationale for TREAT experiment vehicle capabilities. Much of the current progress is focused around understanding and demonstrating the behavior of fuel design with enhanced accident tolerance in water-cooled reactors. Additionally, several related efforts are underway to facilitate transient testing in liquid sodium, inert gas, and steam environments. This paper discusses why such a variety of capabilities are needed, outlines plans to accomplish them, and describes the relationship between transient data needs and the irradiation hardware that will support the gathering of this information.« less

Iron-chrome-aluminum alloy cladding for increasing safety in nuclear power plants

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

2017-12-01

After a tsunami caused plant black out at Fukushima, followed by hydrogen explosions, the US Department of Energy partnered with fuel vendors to study safer alternatives to the current UO2-zirconium alloy system. This accident tolerant fuel alternative should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General electric, Oak ridge national laboratory, and their partners are proposing to replace zirconium alloy cladding in current commercial light water power reactors with an iron-chromium-aluminum (FeCrAl) cladding such as APMT or C26M. Extensive testing and evaluation is being conducted to determine the suitability of FeCrAl under normal operation conditions and under severe accident conditions. Results show that FeCrAl has excellent corrosion resistance under normal operation conditions and FeCrAl is several orders of magnitude more resistant than zirconium alloys to degradation by superheated steam under accident conditions, generating less heat of oxidation and lower amount of combustible hydrogen gas. Higher neutron absorption and tritium release effects can be minimized by design changes. The implementation of FeCrAl cladding is a near term solution to enhance the safety of the current fleet of commercial light water power reactors.

EXPERIMENTAL MOLTEN-SALT-FUELED 30-Mw POWER REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alexander, L.G.; Kinyon, B.W.; Lackey, M.E.

1960-03-24

A preliminary design study was made of an experimental molten-salt- fueled power reactor. The reactor considered is a single-region homogeneous burner coupled with a Loeffler steam-generating cycle. Conceptual plant layouts, basic information on the major fuel circuit components, a process flowsheet, and the nuclear characteristics of the core are presented. The design plant electrical output is 10 Mw, and the total construction cost is estimated to be approximately ,000,000. (auth)

Steam inhalation or humidified oxygen for acute bronchiolitis in children up to three years of age.

PubMed

Umoren, Rachel; Odey, Friday; Meremikwu, Martin M

2011-01-19

Acute bronchiolitis is a common respiratory infection and a major cause of morbidity in young children. It is treated with bronchodilators (for example, salbutamol), corticosteroids or humidified air (steam inhalation or cool mist). Steam inhalation is preferred in low-income countries as it is inexpensive and easily available. It is thought to act as a secretolytic agent to lighten secretions in the respiratory tract and relieve respiratory distress. To evaluate the effect of steam inhalation or humidified oxygen to relieve respiratory distress and to evaluate adverse events in children up to three years old with acute bronchiolitis. We searched the Cochrane Central Register of Controlled Trials (CENTRAL) (The Cochrane Library 2010, Issue 1) which contains the Acute Respiratory Infections Group's Specialised Register, MEDLINE (1950 to February Week 4, 2010), EMBASE.com (1974 to March 2010), CINAHL (1981 to March 2010), AMED (1985 to March 2010), Web of Science (2000 to March 2010) and LILACS (1982 to March 2010). Randomised controlled trials involving children up to three years old with bronchiolitis comparing steam inhalation (or cool mist) or humidified oxygen against bronchodilators, corticosteroids or placebo; alone or in combination. Two review authors independently assessed trial quality and extracted data. Only one study (156 children aged between seven weeks and 24 months with signs and symptoms of bronchiolitis) met the eligibility criteria for inclusion. Participants were randomised into three groups: nebulised salbutamol, nebulised saline and mist in a tent. The results showed a significant decrease in respiratory distress symptom (RDS) score in the nebulised salbutamol group but no significant decrease in the RDS score in the mist in a tent or nebulised saline groups. The study did not report on adverse effects of the interventions. Steam inhalation (or cool mist therapy) is commonly used to treat acute bronchiolitis in resource-constrained settings. One study was eligible for inclusion and found that nebulised salbutamol was an effective intervention for young children with bronchiolitis but mist in a tent did not lead to a significant decrease in RDS score. Since only one study was analysed it would be misleading to conclude that mist therapy is ineffective in children with bronchiolitis. We conclude that there is insufficient evidence to inform practice regarding using steam inhalation or mist therapy for acute bronchiolitis in children up to three years old.

Nonpneumonic, short-incubation-period Legionellosis (Pontiac fever) in men who cleaned a steam turbine condenser.

PubMed

Fraser, D W; Deubner, D C; Hill, D L; Gilliam, D K

1979-08-17

Pontiac fever affected ten men who had cleaned a steam turbine condenser with compressed air. Previous epidemics of Pontiac fever and Legionnaires' disease--both caused by Legionella Pneumophila (proposed sp. nov.)--involved "airborne spread" from air-conditioning cooling towers or evaporative condensers. Aerosols of contaminated water in heat-rejection systems appear to be important sources of epidemic legionellosis.

ADVANCED COURSE ON FUEL ELEMENTS FOR WATER COOLED POWER REACTORS, ORGANIZED BY THE NETHERLANDS'-NORWEGIAN REACTOR SCHOOL AT INSTITUTT FOR ATOMENERGI, KJELLER, NORWAY, 22nd AUGUST-3rd SEPTEMBER,1960. VOLUME III

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aas, S.; Barendregt, T.J.; Chesne, A.

1960-07-01

A series of lectures on fuel elements for water-cooled power reactors are presented. Topics covered include fabrication, properties, cladding, radiation damage, design, cycling, storage and transpont, and reprocessing. Separate records have been prepared for each section.

Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm

NASA Astrophysics Data System (ADS)

Drajat, R. Z.; Su'ud, Z.; Soewono, E.; Gunawan, A. Y.

2012-05-01

There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

Natural circulating passive cooling system for nuclear reactor containment structure

DOEpatents

Gou, Perng-Fei; Wade, Gentry E.

1990-01-01

A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

Molten metal reactor and method of forming hydrogen, carbon monoxide and carbon dioxide using the molten alkaline metal reactor

DOEpatents

Bingham, Dennis N.; Klingler, Kerry M.; Turner, Terry D.; Wilding, Bruce M.

2012-11-13

A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

Method for producing H.sub.2 using a rotating drum reactor with a pulse jet heat source

DOEpatents

Paulson, Leland E.

1990-01-01

A method of producing hydrogen by an endothermic steam-carbon reaction using a rotating drum reactor and a pulse jet combustor. The pulse jet combustor uses coal dust as a fuel to provide reaction temperatures of 1300.degree. to 1400.degree. F. Low-rank coal, water, limestone and catalyst are fed into the drum reactor where they are heated, tumbled and reacted. Part of the reaction product from the rotating drum reactor is hydrogen which can be utilized in suitable devices.

Method of production H/sub 2/ using a rotating drum reactor with a pulse jet heat source

DOEpatents

Paulson, L.E.

1988-05-13

A method of producing hydrogen by an endothermic steam-carbon reaction using a rotating drum reactor and a pulse jet combustor. The pulse jet combustor uses coal dust as a fuel to provide reaction temperatures of 1300/degree/ to 1400/degree/F. Low-rank coal, water, limestone and catalyst are fed into the drum reactor where they are heated, tumbled and reacted. Part of the reaction product from the rotating drum reactor is hydrogen which can be utilized in suitable devices. 1 fig.

Green Chemical Treatments for Heating and Cooling Systems

DTIC Science & Technology

2006-09-01

Legionella pneumophila bacterium, which causes Legion- naire’s Disease. 2.3 Steam Line Treatment The third and final product in the Green Chemistry...ER D C/ CE R L TR -0 6 -2 9 Green Chemical Treatments for Heating and Cooling Systems Susan A. Drozdz and Vincent F. Hock September...CERL TR-06-29 September 2006 Green Chemical Treatments for Heating and Cooling Systems Susan A. Drozdz and Vincent F. Hock Construction

Economic analysis of condensers for water recovery in steam injected gas turbines

DOE Office of Scientific and Technical Information (OSTI.GOV)

De Paepe, M.; Huvenne, P.; Dick, E.

1998-07-01

Steam injection cycles are interesting for small power ranges because of the high efficiency and the relatively low investment costs. A big disadvantage is the consumption of water by the cycle. Water recovery is seldom realized in industrial practice. In this paper an analysis of the technical and economical possibilities of water recovery by condensation of water out of the exhaust gases is made. Three gas turbines are considered : the Kawasaki M1A-13CC (2.3 MWe), the Allison 501KH (6.8 MWe) and the General Electric LM1600 (17 MWe). For every gas turbine two types of condensers are designed. In the watermore » cooled condenser finned tubes are used to cool the exhaust gases, flowing at the outside of the tubes. The water itself flows at the inside of the tubes and is cooled by a water to air cooler. In the air cooled condenser the exhaust gases flow at the inside of the tubes and the cooling air at the outside. The investment costs of the condensers is compared to the costs of the total installation. The investment costs are relatively smaller if the produced power goes up. The water cooled condenser with water to air cooler is cheaper than the air cooled condenser. Using a condenser results in higher exploitation costs due to the fans and pumps. It is shown that the air cooled condenser has lower exploitation costs than the water cooled one. Pay back time of the total installation does not significantly vary compared to the installation without recovery. Water prices are determined for which water recovery is profitable. For the water cooled condenser the turning point lies at 2.2 Euro/m; for the air cooled condenser this is 0.6 Euro/m.« less

Water chemistry of the secondary circuit at a nuclear power station with a VVER power reactor

NASA Astrophysics Data System (ADS)

Tyapkov, V. F.; Erpyleva, S. F.

2017-05-01

Results of implementation of the secondary circuit organic amine water chemistry at Russian nuclear power plant (NPP) with VVER-1000 reactors are presented. The requirements for improving the reliability, safety, and efficiency of NPPs and for prolonging the service life of main equipment items necessitate the implementation of new technologies, such as new water chemistries. Data are analyzed on the chemical control of power unit coolant for quality after the changeover to operation with the feed of higher amines, such as morpholine and ethanolamine. Power units having equipment containing copper alloy components were converted from the all-volatile water chemistry to the ethanolamine or morpholine water chemistry with no increase in pH of the steam generator feedwater. This enables the iron content in the steam generator feedwater to be decreased from 6-12 to 2.0-2.5 μg/dm3. It is demonstrated that pH of high-temperature water is among the basic factors controlling erosion and corrosion wear of the piping and the ingress of corrosion products into NPP steam generators. For NPP power units having equipment whose construction material does not include copper alloys, the water chemistries with elevated pH of the secondary coolant are adopted. Stable dosing of correction chemicals at these power units maintains pH25 of 9.5 to 9.7 in the steam generator feedwater with a maximum iron content of 2 μg/dm3 in the steam generator feedwater.

Gasification system

DOEpatents

Haldipur, Gaurang B.; Anderson, Richard G.; Cherish, Peter

1985-01-01

A method and system for injecting coal and process fluids into a fluidized bed gasification reactor. Three concentric tubes extend vertically upward into the fluidized bed. Coal particulates in a transport gas are injected through an inner tube, and an oxygen rich mixture of oxygen and steam are injected through an inner annulus about the inner tube. A gaseous medium relatively lean in oxygen content, such as steam, is injected through an annulus surrounding the inner annulus.

Gasification system

DOEpatents

Haldipur, Gaurang B.; Anderson, Richard G.; Cherish, Peter

1983-01-01

A method and system for injecting coal and process fluids into a fluidized bed gasification reactor. Three concentric tubes extend vertically upward into the fluidized bed. Coal particulates in a transport gas are injected through an inner tube, and an oxygen rich mixture of oxygen and steam are injected through an inner annulus about the inner tube. A gaseous medium relatively lean in oxygen content, such as steam, is injected through an annulus surrounding the inner annulus.

Development concept for a small, split-core, heat-pipe-cooled nuclear reactor

NASA Technical Reports Server (NTRS)

Lantz, E.; Breitwieser, R.; Niederauer, G. F.

1974-01-01

There have been two main deterrents to the development of semiportable nuclear reactors. One is the high development costs; the other is the inability to satisfy with assurance the questions of operational safety. This report shows how a split-core, heat-pipe cooled reactor could conceptually eliminate these deterrents, and examines and summarizes recent work on split-core, heat-pipe reactors. A concept for a small reactor that could be developed at a comparatively low cost is presented. The concept would extend the technology of subcritical radioisotope thermoelectric generators using 238 PuO2 to the evolution of critical space power reactors using 239 PuO2.

A numerical analysis of heat and mass transfer during the steam reforming process of ethane

NASA Astrophysics Data System (ADS)

Tomiczek, Marcin; Kaczmarczyk, Robert; Mozdzierz, Marcin; Brus, Grzegorz

2017-11-01

This paper presents a numerical analysis of heat and mass transfer during the steam reforming of ethane. From a chemical point of view, the reforming process of heavy hydrocarbons, such as ethane, is complex. One of the main issue is a set of undesired chemical reactions that causes the deposition of solid carbon and consequently blocks the catalytic property of a reactor. In the literature a carbon deposition regime is selected by thermodynamical analysis to design safe operation conditions. In the case of Computational Fluid Dynamic (CFD, hereafter) models each control volume should be investigated to determinate if carbon deposition is thermodynamically favourable. In this paper the authors combine equilibrium and kinetics analysis to simulate the steam reforming of methane-ethane rich fuel. The results of the computations were juxtaposed with experimental data for methane steam reforming, and good agreement was found. An analysis based on the kinetics of reactions was conducted to predict the influence of temperature drop and non-equilibrium composition on solid carbon deposition. It was found that strong non-uniform temperature distribution in the reactor causes conditions favourable for carbon deposition at the inlet of the reformer. It was shown that equilibrium calculations, often used in the literature, are insufficient.

Sorption enhanced reaction process (SERP) for the production of hydrogen

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hufton, J.; Mayorga, S.; Gaffney, T.

1998-08-01

The novel Sorption Enhanced Reaction Process has the potential to decrease the cost of hydrogen production by steam methane reforming. Current effort for development of this technology has focused on adsorbent development, experimental process concept testing, and process development and design. A preferred CO{sub 2} adsorbent, K{sub 2}CO{sub 3} promoted hydrotalcite, satisfies all of the performance targets and it has been scaled up for process testing. A separate class of adsorbents has been identified which could potentially improve the performance of the H{sub 2}-SER process. Although this material exhibits improved CO{sub 2} adsorption capacity compared to the HTC adsorbent, itsmore » hydrothermal stability must be improved. Single-step process experiments (not cyclic) indicate that the H{sub 2}-SER reactor performance during the reaction step improves with decreasing pressure and increasing temperature and steam to methane ratio in the feed. Methane conversion in the H{sub 2}-SER reactor is higher than for a conventional catalyst-only reactor operated at similar temperature and pressure. The reactor effluent gas consists of 90+% H{sub 2}, balance CH{sub 4}, with only trace levels (< 50 ppm) of carbon oxides. A best-case process design (2.5 MMSCFD of 99.9+% H{sub 2}) based on the HTC adsorbent properties and a revised SER process cycle has been generated. Economic analysis of this design indicates the process has the potential to reduce the H{sub 2} product cost by 25--31% compared to conventional steam methane reforming.« less

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.

1983-06-01

During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

Passive heat-transfer means for nuclear reactors. [LMFBR

DOEpatents

Burelbach, J.P.

1982-06-10

An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

Summary Report On Design And Development Of High Temperature Gas-Cooled Power Pile

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCullough, C. R.

1947-09-15

This report presents a description of a design for an experimental nuclear power plant utilizing a high temperature gas-cooled power pile as the energy source. The plant consists of the pile, a heat exchanger or boiler, a conventional steam turbine generator and their associated auxiliaries. Helium gas under pressure transfers heat from the pile to the boiler which generates steam for driving the generator. The plant is rated at a normal output of 12,000 kilowatts of heat and an electrical output of 2400 kilowatts. Provision is made for operation up to 20,000 kilowatts of heat (4000 kilowatts of electrical output)more » in the event operation of the plants proves this possible.« less

77 FR 64148 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Regulatory...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-18

... Regulatory Guides (RG) RG 1.79, ````Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors,'' Revision 2 and RG 1.79.1, ``Initial Test Program of Emergency Core Cooling Systems for...

DETECTION OF COATING FAILURES IN A NEUTRONIC REACTOR

DOEpatents

Snell, A.H.; Allison, S.K.

1958-02-11

This patent relates to water-cooled reactor systems and discloses a means to detect leaks in the jackets of jacketed fuel elements comprising a neutron detector located in the cooling water discharge pipe,the pipe being provided with an enlarged portion for housing the detector so that the latter is completely surrounded by the water in its passage through the pipe, said enlarged portion and detector being shielded from the reactor for the purpose of detecting only those delayed neutrons emitted in the cooling water and due to the latter picking up fission fragments from the defective fuel elements.

High-Temperature Gas-Cooled Test Reactor Point Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

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.

Numerical Simulation and Analyses of the Loss of Feedwater Transient at the Unit 4 of Kola NPP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stevanovic, Vladimir D.; Stosic, Zoran V.; Kiera, Michael

2002-07-01

A three-dimensional numerical simulation of the loss-of-feed water transient at the horizontal steam generator of the Kola nuclear power plant is performed. Presented numerical results show transient change of integral steam generator parameters, such as steam generation rate, water mass inventory, outlet reactor coolant temperature, as well as detailed distribution of shell side thermal-hydraulic parameters: swell and collapsed levels, void fraction distributions, mass flux vectors, etc. Numerical results are compared with measurements at the Kola NPP. The agreement is satisfactory, while differences are close to or below the measurement uncertainties. Obtained numerical results are the first ones that give completemore » insight into the three-dimensional and transient horizontal steam generator thermal-hydraulics. Also, the presented results serve as benchmark tests for the assessment and further improvement of one-dimensional models of horizontal steam generator built with safety codes. (authors)« less

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2013 CFR

2013-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power... light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter. The...

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2012 CFR

2012-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power... light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter. The...

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2014 CFR

2014-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power... light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter. The...

The international water conference proceedings

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guseman, J.R.

1984-10-01

This book provides information on computer applications to water chemistry control, groundwater, membrane technology, instrumentation/analytical techniques and ion exchange. Other topics of discussion include cooling water, biocontrol, the hydraulic properties of ion exchange resins, steam electric power plant aqueous discharges and colorimetric determination of trace benzotriazole or tolytriazole. Water chemistry guidelines for large steam generating power plants is discussed, as well as wastewater treatment, boiler water conditioning and ion exchange/computer related topics.

Fuel processing for PEM fuel cells: transport and kinetic issues of system design

NASA Astrophysics Data System (ADS)

Zalc, J. M.; Löffler, D. G.

In light of the distribution and storage issues associated with hydrogen, efficient on-board fuel processing will be a significant factor in the implementation of PEM fuel cells for automotive applications. Here, we apply basic chemical engineering principles to gain insight into the factors that limit performance in each component of a fuel processor. A system consisting of a plate reactor steam reformer, water-gas shift unit, and preferential oxidation reactor is used as a case study. It is found that for a steam reformer based on catalyst-coated foils, mass transfer from the bulk gas to the catalyst surface is the limiting process. The water-gas shift reactor is expected to be the largest component of the fuel processor and is limited by intrinsic catalyst activity, while a successful preferential oxidation unit depends on strict temperature control in order to minimize parasitic hydrogen oxidation. This stepwise approach of sequentially eliminating rate-limiting processes can be used to identify possible means of performance enhancement in a broad range of applications.

Performance of a natural gas fuel processor for residential PEFC system using a novel CO preferential oxidation catalyst

NASA Astrophysics Data System (ADS)

Echigo, Mitsuaki; Shinke, Norihisa; Takami, Susumu; Tabata, Takeshi

Natural gas fuel processors have been developed for 500 W and 1 kW class residential polymer electrolyte fuel cell (PEFC) systems. These fuel processors contain all the elements—desulfurizers, steam reformers, CO shift converters, CO preferential oxidation (PROX) reactors, steam generators, burners and heat exchangers—in one package. For the PROX reactor, a single-stage PROX process using a novel PROX catalyst was adopted. In the 1 kW class fuel processor, thermal efficiency of 83% at HHV was achieved at nominal output assuming a H 2 utilization rate in the cell stack of 76%. CO concentration below 1 ppm in the product gas was achieved even under the condition of [O 2]/[CO]=1.5 at the PROX reactor. The long-term durability of the fuel processor was demonstrated with almost no deterioration in thermal efficiency and CO concentration for 10,000 h, 1000 times start and stop cycles, 25,000 cycles of load change.

Passive containment cooling system

DOEpatents

Conway, Lawrence E.; Stewart, William A.

1991-01-01

A containment cooling system utilizes a naturally induced air flow and a gravity flow of water over the containment shell which encloses a reactor core to cool reactor core decay heat in two stages. When core decay heat is greatest, the water and air flow combine to provide adequate evaporative cooling as heat from within the containment is transferred to the water flowing over the same. The water is heated by heat transfer and then evaporated and removed by the air flow. After an initial period of about three to four days when core decay heat is greatest, air flow alone is sufficient to cool the containment.

From Steaming Mad to Staying Cool: A Constructive Approach to Anger Control

ERIC Educational Resources Information Center

Feindler, Eva L.; Starr, Karen E.

2003-01-01

Teaching children and adolescents to recognize how they feel when they are angry and what pushes their buttons enables them to make better choices about how they express their anger. They learn that staying cool gives them the power to create more positive outcomes for potentially negative encounters. Through self-assessment and role-plays, they…

Combined Steady-State and Dynamic Heat Exchanger Experiment

ERIC Educational Resources Information Center

Luyben, William L.; Tuzla, Kemal; Bader, Paul N.

2009-01-01

This paper describes a heat-transfer experiment that combines steady-state analysis and dynamic control. A process-water stream is circulated through two tube-in-shell heat exchangers in series. In the first, the process water is heated by steam. In the second, it is cooled by cooling water. The equipment is pilot-plant size: heat-transfer areas…

Gasification of carbonaceous solids

DOEpatents

Coates, Ralph L.

1976-10-26

A process and apparatus for converting coal and other carbonaceous solids to an intermediate heating value fuel gas or to a synthesis gas. A stream of entrained pulverized coal is fed into the combustion stage of a three-stage gasifier along with a mixture of oxygen and steam at selected pressure and temperature. The products of the combustion stage pass into the second or quench stage where they are partially cooled and further reacted with water and/or steam. Ash is solidified into small particles and the formation of soot is suppressed by water/steam injections in the quench stage. The design of the quench stage prevents slag from solidifying on the walls. The products from the quench stage pass directly into a heat recovery stage where the products pass through the tube, or tubes, of a single-pass, shell and tube heat exchanger and steam is generated on the shell side and utilized for steam feed requirements of the process.

Condensation induced water hammer driven sterilization

DOEpatents

Kullberg, Craig M.

2004-05-11

A method and apparatus (10) for treating a fluid or materials therein with acoustic energy has a vessel (14) for receiving the fluid with inner walls shaped to focus acoustic energy to a target zone within the vessel. One or more nozzles (26) are directed into the vessel (14) for injecting a condensable vapor, such as steam, into the vessel (14). The system may include a steam source (18) for providing steam as the condensable vapor from an industrial waste heat source. Steam drums (88) are disposed between the steam source (18) and nozzles (26) to equalize and distribute the vapor pressure. A cooling source (30) provides a secondary fluid for maintaining the liquid in the vessel (14) in subcooled conditions. A heating jacket (32) surrounds the vessel (14) to heat the walls of the vessel (14) and prevent biological growth thereon. A pressurizer (33) may operate the system at elevated pressures.

Fuel-Cell Power Systems Incorporating Mg-Based H2 Generators

NASA Technical Reports Server (NTRS)

Kindler, Andrew; Narayan, Sri R.

2009-01-01

Two hydrogen generators based on reactions involving magnesium and steam have been proposed as means for generating the fuel (hydrogen gas) for such fuel-cell power systems as those to be used in the drive systems of advanced motor vehicles. The hydrogen generators would make it unnecessary to rely on any of the hydrogen storage systems developed thus far that are, variously, too expensive, too heavy, too bulky, and/or too unsafe to be practical. The two proposed hydrogen generators are denoted basic and advanced, respectively. In the basic hydrogen generator (see figure), steam at a temperature greater than or equals 330 C would be fed into a reactor charged with magnesium, wherein hydrogen would be released in the exothermic reaction Mg + H2O yields MgO + H2. The steam would be made in a flash boiler. To initiate the reaction, the boiler could be heated electrically by energy borrowed from a storage battery that would be recharged during normal operation of the associated fuel-cell subsystem. Once the reaction was underway, heat from the reaction would be fed to the boiler. If the boiler were made an integral part of the hydrogen-generator reactor vessel, then the problem of transfer of heat from the reactor to the boiler would be greatly simplified. A pump would be used to feed water from a storage tank to the boiler.

Open cycle ocean thermal energy conversion system structure

DOEpatents

Wittig, J. Michael

1980-01-01

A generally mushroom-shaped, open cycle OTEC system and distilled water producer which has a skirt-conduit structure extending from the enlarged portion of the mushroom to the ocean. The enlarged part of the mushroom houses a toroidal casing flash evaporator which produces steam which expands through a vertical rotor turbine, partially situated in the center of the blossom portion and partially situated in the mushroom's stem portion. Upon expansion through the turbine, the motive steam enters a shell and tube condenser annularly disposed about the rotor axis and axially situated beneath the turbine in the stem portion. Relatively warm ocean water is circulated up through the radially outer skirt-conduit structure entering the evaporator through a radially outer portion thereof, flashing a portion thereof into motive steam, and draining the unflashed portion from the evaporator through a radially inner skirt-conduit structure. Relatively cold cooling water enters the annular condenser through the radially inner edge and travels radially outwardly into a channel situated along the radially outer edge of the condenser. The channel is also included in the radially inner skirt-conduit structure. The cooling water is segregated from the potable, motive steam condensate which can be used for human consumption or other processes requiring high purity water. The expansion energy of the motive steam is partially converted into rotational mechanical energy of the turbine rotor when the steam is expanded through the shaft attached blades. Such mechanical energy drives a generator also included in the enlarged mushroom portion for producing electrical energy. Such power generation equipment arrangement provides a compact power system from which additional benefits may be obtained by fabricating the enclosing equipment, housings and component casings from low density materials, such as prestressed concrete, to permit those casings and housings to also function as a floating support vessel.

Heat transfer during condensation of steam from steam-gas mixtures in the passive safety systems of nuclear power plants

NASA Astrophysics Data System (ADS)

Portnova, N. M.; Smirnov, Yu B.

2017-11-01

A theoretical model for calculation of heat transfer during condensation of multicomponent vapor-gas mixtures on vertical surfaces, based on film theory and heat and mass transfer analogy is proposed. Calculations were performed for the conditions implemented in experimental studies of heat transfer during condensation of steam-gas mixtures in the passive safety systems of PWR-type reactors of different designs. Calculated values of heat transfer coefficients for condensation of steam-air, steam-air-helium and steam-air-hydrogen mixtures at pressures of 0.2 to 0.6 MPa and of steam-nitrogen mixture at the pressures of 0.4 to 2.6 MPa were obtained. The composition of mixtures and vapor-to-surface temperature difference were varied within wide limits. Tube length ranged from 0.65 to 9.79m. The condensation of all steam-gas mixtures took place in a laminar-wave flow mode of condensate film and turbulent free convection in the diffusion boundary layer. The heat transfer coefficients obtained by calculation using the proposed model are in good agreement with the considered experimental data for both the binary and ternary mixtures.

Dynamic modeling of temperature change in outdoor operated tubular photobioreactors.

PubMed

Androga, Dominic Deo; Uyar, Basar; Koku, Harun; Eroglu, Inci

2017-07-01

In this study, a one-dimensional transient model was developed to analyze the temperature variation of tubular photobioreactors operated outdoors and the validity of the model was tested by comparing the predictions of the model with the experimental data. The model included the effects of convection and radiative heat exchange on the reactor temperature throughout the day. The temperatures in the reactors increased with increasing solar radiation and air temperatures, and the predicted reactor temperatures corresponded well to the measured experimental values. The heat transferred to the reactor was mainly through radiation: the radiative heat absorbed by the reactor medium, ground radiation, air radiation, and solar (direct and diffuse) radiation, while heat loss was mainly through the heat transfer to the cooling water and forced convection. The amount of heat transferred by reflected radiation and metabolic activities of the bacteria and pump work was negligible. Counter-current cooling was more effective in controlling reactor temperature than co-current cooling. The model developed identifies major heat transfer mechanisms in outdoor operated tubular photobioreactors, and accurately predicts temperature changes in these systems. This is useful in determining cooling duty under transient conditions and scaling up photobioreactors. The photobioreactor design and the thermal modeling were carried out and experimental results obtained for the case study of photofermentative hydrogen production by Rhodobacter capsulatus, but the approach is applicable to photobiological systems that are to be operated under outdoor conditions with significant cooling demands.

Optimization of power-cycle arrangements for Supercritical Water cooled Reactors (SCWRs)

NASA Astrophysics Data System (ADS)

Lizon-A-Lugrin, Laure

The world energy demand is continuously rising due to the increase of both the world population and the standard of life quality. Further, to assure both a healthy world economy as well as adequate social standards, in a relatively short term, new energy-conversion technologies are mandatory. Within this framework, a Generation IV International Forum (GIF) was established by the participation of 10 countries to collaborate for developing nuclear power reactors that will replace the present technology by 2030. The main goals of these nuclear-power reactors are: economic competitiveness, sustainability, safety, reliability and resistance to proliferation. As a member of the GIF, Canada has decided to orient its efforts towards the design of a CANDU-type Super Critical Water-cooled Reactor (SCWR). Such a system must run at a coolant outlet temperature of about 625°C and at a pressure of 25 MPa. It is obvious that at such conditions the overall efficiency of this kind of Nuclear Power Plant (NPP) will compete with actual supercritical water-power boilers. In addition, from a heat-transfer viewpoint, the use of a supercritical fluid allows the limitation imposed by Critical Heat Flux (CHF) conditions, which characterize actual technologies, to be removed. Furthermore, it will be also possible to use direct thermodynamic cycles where the supercritical fluid expands right away in a turbine without the necessity of using intermediate steam generators and/or separators. This work presents several thermodynamic cycles that could be appropriate to run SCWR power plants. Improving both thermal efficiency and mechanical power constitutes a multi-objective optimization problem and requires specific tools. To this aim, an efficient and robust evolutionary algorithm, based on genetic algorithm, is used and coupled to an appropriate power plant thermodynamic simulation model. The results provide numerous combinations to achieve a thermal efficiency higher than 50% with a mechanical power of 1200 MW. It is observed that in most cases the landscape of Pareto's front is mostly controlled only by few key parameters. These results may be very useful for future plant design engineers. Furthermore, some calculations for pipe sizing and temperature variation between coolant and fuel have been carried out to provide an idea on their order of magnitude.

Preliminary design of high temperature ultrasonic transducers for liquid sodium environments

NASA Astrophysics Data System (ADS)

Prowant, M. S.; Dib, G.; Qiao, H.; Good, M. S.; Larche, M. R.; Sexton, S. S.; Ramuhalli, P.

2018-04-01

Advanced reactor concepts include fast reactors (including sodium-cooled fast reactors), gas-cooled reactors, and molten-salt reactors. Common to these concepts is a higher operating temperature (when compared to light-water-cooled reactors), and the proposed use of new alloys with which there is limited operational experience. Concerns about new degradation mechanisms, such as high-temperature creep and creep fatigue, that are not encountered in the light-water fleet and longer operating cycles between refueling intervals indicate the need for condition monitoring technology. Specific needs in this context include periodic in-service inspection technology for the detection and sizing of cracking, as well as technologies for continuous monitoring of components using in situ probes. This paper will discuss research on the development and evaluation of high temperature (>550°C; >1022°F) ultrasonic probes that can be used for continuous monitoring of components. The focus of this work is on probes that are compatible with a liquid sodium-cooled reactor environment, where the core outlet temperatures can reach 550°C (1022°F). Modeling to assess sensitivity of various sensor configurations and experimental evaluation have pointed to a preferred design and concept of operations for these probes. This paper will describe these studies and ongoing work to fabricate and fully evaluate survivability and sensor performance over extended periods at operational temperatures.

Breckinridge Project, initial effort

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1982-01-01

The project cogeneration plant supplies electric power, process steam and treated boiler feedwater for use by the project plants. The plant consists of multiple turbine generators and steam generators connected to a common main steam header. The major plant systems which are required to produce steam, electrical power and treated feedwater are discussed individually. The systems are: steam, steam generator, steam generator fuel, condensate and feedwater deaeration, condensate and blowdown collection, cooling water, boiler feedwater treatment, coal handling, ash handling (fly ash and bottom ash), electrical, and control system. The plant description is based on the Phase Zero design basismore » established for Plant 31 in July of 1980 and the steam/condensate balance as presented on Drawing 31-E-B-1. Updating of steam requirements as more refined process information becomes available has generated some changes in the steam balance. Boiler operation with these updated requirements is reflected on Drawing 31-D-B-1A. The major impact of updating has been that less 600 psig steam generated within the process units requires more extraction steam from the turbine generators to close the 600 psig steam balance. Since the 900 psig steam generation from the boilers was fixed at 1,200,000 lb/hr, the additional extraction steam required to close the 600 psig steam balance decreased the quantity of electrical power available from the turbine generators. In the next phase of engineering work, the production of 600 psig steam will be augmented by increasing convection bank steam generation in the Plant 3 fired heaters by 140,000 to 150,000 lb/hr. This modification will allow full rated power generation from the turbine generators.« less

Heat dissipating nuclear reactor with metal liner

DOEpatents

Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

1985-11-21

A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

Heat dissipating nuclear reactor with metal liner

DOEpatents

Gluekler, Emil L.; Hunsbedt, Anstein; Lazarus, Jonathan D.

1987-01-01

Disclosed is a nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

A Gas-Cooled-Reactor Closed-Brayton-Cycle Demonstration with Nuclear Heating

NASA Astrophysics Data System (ADS)

Lipinski, Ronald J.; Wright, Steven A.; Dorsey, Daniel J.; Peters, Curtis D.; Brown, Nicholas; Williamson, Joshua; Jablonski, Jennifer

2005-02-01

A gas-cooled reactor may be coupled directly to turbomachinery to form a closed-Brayton-cycle (CBC) system in which the CBC working fluid serves as the reactor coolant. Such a system has the potential to be a very simple and robust space-reactor power system. Gas-cooled reactors have been built and operated in the past, but very few have been coupled directly to the turbomachinery in this fashion. In this paper we describe the option for testing such a system with a small reactor and turbomachinery at Sandia National Laboratories. Sandia currently operates the Annular Core Research Reactor (ACRR) at steady-state powers up to 4 MW and has an adjacent facility with heavy shielding in which another reactor recently operated. Sandia also has a closed-Brayton-Cycle test bed with a converted commercial turbomachinery unit that is rated for up to 30 kWe of power. It is proposed to construct a small experimental gas-cooled reactor core and attach this via ducting to the CBC turbomachinery for cooling and electricity production. Calculations suggest that such a unit could produce about 20 kWe, which would be a good power level for initial surface power units on the Moon or Mars. The intent of this experiment is to demonstrate the stable start-up and operation of such a system. Of particular interest is the effect of a negative temperature power coefficient as the initially cold Brayton gas passes through the core during startup or power changes. Sandia's dynamic model for such a system would be compared with the performance data. This paper describes the neutronics, heat transfer, and cycle dynamics of this proposed system. Safety and radiation issues are presented. The views expressed in this document are those of the author and do not necessarily reflect agreement by the government.

Probability of in-vessel steam explosion-induced containment failure for a KWU PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Esmaili, H.; Khatib-Rahbar, M.; Zuchuat, O.

During postulated core meltdown accidents in light water reactors, there is a likelihood for an in-vessel steam explosion when the melt contacts the coolant in the lower plenum. The objective of the work described in this paper is to determine the conditional probability of in-vessel steam explosion-induced containment failure for a Kraftwerk Union (KWU) pressurized water reactor (PWR). The energetics of the explosion depends on the mass of the molten fuel that mixes with the coolant and participates in the explosion and on the conversion of fuel thermal energy into mechanical work. The work can result in the generation ofmore » dynamic pressures that affect the lower head (and possibly lead to its failure), and it can cause acceleration of a slug (fuel and coolant material) upward that can affect the upper internal structures and vessel head and ultimately cause the failure of the upper head. If the upper head missile has sufficient energy, it can reach the containment shell and penetrate it. The analysis, must therefore, take into account all possible dissipation mechanisms.« less

An afterburner-powered methane/steam reformer for a solid oxide fuel cells application

NASA Astrophysics Data System (ADS)

Mozdzierz, Marcin; Chalusiak, Maciej; Kimijima, Shinji; Szmyd, Janusz S.; Brus, Grzegorz

2018-04-01

Solid oxide fuel cell (SOFC) systems can be fueled by natural gas when the reforming reaction is conducted in a stack. Due to its maturity and safety, indirect internal reforming is usually used. A strong endothermic methane/steam reforming process needs a large amount of heat, and it is convenient to provide thermal energy by burning the remainders of fuel from a cell. In this work, the mathematical model of afterburner-powered methane/steam reformer is proposed. To analyze the effect of a fuel composition on SOFC performance, the zero-dimensional model of a fuel cell connected with a reformer is formulated. It is shown that the highest efficiency of a solid oxide fuel cell is achieved when the steam-to-methane ratio at the reforming reactor inlet is high.

Process And Apparatus To Accomplish Autothermal Or Steam Reforming Via A Reciprocating Compression Device

DOEpatents

Lyons, K. David; James, Robert; Berry, David A.; Gardner, Todd

2004-09-21

The invention provides a method and apparatus for producing a synthesis gas from a variety of hydrocarbons. The apparatus (device) consists of a semi-batch, non-constant volume reactor to generate a synthesis gas. While the apparatus feeds mixtures of air, steam, and hydrocarbons into a cylinder where work is performed on the fluid by a piston to adiabatically raise its temperature without heat transfer from an external source.

Morphological changes in the cellulose and lignin components of biomass occur at different stages of steam pretreatment

DOE PAGES

Pingali, Sai Venkatesh; O'Neill, Hugh Michael; Nishiyama, Yoshiharu; ...

2014-01-09

Morphological changes to the different components of lignocellulosic biomass were observed as they occurred during steam pretreatment by placing a pressure reaction cell in a neutron beam and collecting time-resolved neutron scattering data. Changes to cellulose morphology occurred mainly in the heating phase, whereas changes in lignin morphology occurred mainly in the holding and cooling phases. During the heating stage, water is irreversibly expelled from cellulose microfibrils as the elementary fibrils coalesce. During the holding phase lignin aggregates begin to appear and they increase in size most noticeably during the cooling phase. This experiment demonstrates the unique information that inmore » situ small angle neutron scattering studies of pretreatment can provide. This approach is potentially useful in optimizing the heating, holding and cooling stages of pretreatments to allow the exact size and nature of lignin aggregates to be controlled in order to enhance enzyme accessibility to cellulose and therefore the efficiency of biomass conversion.« less

Oxidation Kinetics of Ferritic Alloys in High-Temperature Steam Environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parker, Stephen S.; White, Josh; Hosemann, Peter

High-temperature isothermal steam oxidation kinetic parameters of several ferritic alloys were determined by thermogravimetric analysis. We measured the oxidation kinetic constant (k) as a function of temperature from 900°C to 1200°C. The results show a marked increase in oxidation resistance compared to reference Zircaloy-2, with kinetic constants 3–5 orders of magnitude lower across the experimental temperature range. Our results of this investigation supplement previous findings on the properties of ferritic alloys for use as candidate cladding materials and extend kinetic parameter measurements to high-temperature steam environments suitable for assessing accident tolerance for light water reactor applications.

Oxidation Kinetics of Ferritic Alloys in High-Temperature Steam Environments

DOE PAGES

Parker, Stephen S.; White, Josh; Hosemann, Peter; ...

2017-11-03

High-temperature isothermal steam oxidation kinetic parameters of several ferritic alloys were determined by thermogravimetric analysis. We measured the oxidation kinetic constant (k) as a function of temperature from 900°C to 1200°C. The results show a marked increase in oxidation resistance compared to reference Zircaloy-2, with kinetic constants 3–5 orders of magnitude lower across the experimental temperature range. Our results of this investigation supplement previous findings on the properties of ferritic alloys for use as candidate cladding materials and extend kinetic parameter measurements to high-temperature steam environments suitable for assessing accident tolerance for light water reactor applications.

Oxidation Kinetics of Ferritic Alloys in High-Temperature Steam Environments

NASA Astrophysics Data System (ADS)

Parker, Stephen S.; White, Josh; Hosemann, Peter; Nelson, Andrew

2018-02-01

High-temperature isothermal steam oxidation kinetic parameters of several ferritic alloys were determined by thermogravimetric analysis. The oxidation kinetic constant ( k) was measured as a function of temperature from 900°C to 1200°C. The results show a marked increase in oxidation resistance compared to reference Zircaloy-2, with kinetic constants 3-5 orders of magnitude lower across the experimental temperature range. The results of this investigation supplement previous findings on the properties of ferritic alloys for use as candidate cladding materials and extend kinetic parameter measurements to high-temperature steam environments suitable for assessing accident tolerance for light water reactor applications.

Erosion-corrosion entrainment of iron-containing compounds as a source of deposits in steam generators used at nuclear power plants equipped with VVER reactors

NASA Astrophysics Data System (ADS)

Tomarov, G. V.; Shipkov, A. A.

2011-03-01

The main stages and processes through which deposits are generated, migrate, and precipitate in the metal-secondary coolant system of power units at nuclear power plants are analyzed and determined. It is shown that substances produced by the mechanism of general erosion-corrosion are the main source of the ionic-colloid form of iron, which is the main component of deposits in a steam generator. Ways for controlling the formation of deposits in a nuclear power plant's steam generator are proposed together with methods for estimating their efficiency.

A thermodynamic approach for advanced fuels of gas-cooled reactors

NASA Astrophysics Data System (ADS)

Guéneau, C.; Chatain, S.; Gossé, S.; Rado, C.; Rapaud, O.; Lechelle, J.; Dumas, J. C.; Chatillon, C.

2005-09-01

For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO 2 gas formation during the chemical interaction of [UO 2± x/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

Katie Gaston | NREL

Science.gov Websites

gasification. The new catalytic riser reactor system can be used for ex situ vapor-phase upgrading or steam , and Wood: 1. Parametric Study and Comparison with Literature," Industrial & Engineering

Oxygen transport membrane based advanced power cycle with low pressure synthesis gas slip stream

DOEpatents

Kromer, Brian R.; Litwin, Michael M.; Kelly, Sean M.

2016-09-27

A method and system for generating electrical power in which a high pressure synthesis gas stream generated in a gasifier is partially oxidized in an oxygen transport membrane based reactor, expanded and thereafter, is combusted in an oxygen transport membrane based boiler. A low pressure synthesis gas slip stream is split off downstream of the expanders and used as the source of fuel in the oxygen transport membrane based partial oxidation reactors to allow the oxygen transport membrane to operate at low fuel pressures with high fuel utilization. The combustion within the boiler generates heat to raise steam to in turn generate electricity by a generator coupled to a steam turbine. The resultant flue gas can be purified to produce a carbon dioxide product.

Development of High-Powered Steam Turbines by OAO NPO Central Research and Design Institute for Boilers and Turbines

NASA Astrophysics Data System (ADS)

Mikhailov, V. E.; Khomenok, L. A.; Kovalev, I. A.

2018-01-01

The article provides an overview of the developments by OAO NPO TsKTI aimed at improvement of components and assemblies of new-generation turbine plants for ultra-supercritical steam parameters to be installed at the power-generating facilities in service. The list of the assemblies under development includes cylinder shells, the cylinder's flow paths and rotors, seals, bearings, and rotor cooling systems. The authors consider variants of the shafting-cylinder configurations for which advanced high-pressure and intermediate-pressure cylinders with reactive blading and low-pressure cylinders of conventional design and with counter-current steam flows are proposed and high-pressure rotors, which can increase the economic efficiency and reduce the overall turbine plant dimensions. Materials intended for the equipment components that operate at high temperatures and a steam cooling technique that allows the use of cheaper steel grades owing to the reduction in the metal's working temperature are proposed. A new promising material for the bearing surfaces is described that enables the operation at higher unit pressures. The material was tested on a full-scale test bench at OAO NPO TsKTI and a turbine in operation. Ways of controlling the erosion of the blades in the moisture-steam turbine compartments by the steam heating of the hollow guide blades are considered. To ensure the dynamic stability of the shafting, shroud and diaphragm seals that prevent the development of the destabilizing circulatory forces of the steam flow were devised and trialed. Advanced instrumentation and software are proposed to monitor the condition of the blading and thermal stresses under transient conditions, to diagnose the vibration processes, and to archive the obtained data. Attention is paid to the normalization of the electromagnetic state of the plant in order to prevent the electrolytic erosion of the plant components. The instrumentation intended for monitoring the relevant electric parameters is described.

Monitoring system for a liquid-cooled nuclear fission reactor. [PWR

DOEpatents

DeVolpi, A.

1984-07-20

The invention provides improved means for detecting the water levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting the density of the water in these regions. The invention utilizes a plurality of exterior gamma radiation detectors and a collimator technique operable to sense separate regions of the reactor vessel to give respectively, unique signals for these regions, whereby comparative analysis of these signals can be used to advise of the presence and density of cooling water in the vessel.

Optimal design of solid oxide fuel cell, ammonia-water single effect absorption cycle and Rankine steam cycle hybrid system

NASA Astrophysics Data System (ADS)

Mehrpooya, Mehdi; Dehghani, Hossein; Ali Moosavian, S. M.

2016-02-01

A combined system containing solid oxide fuel cell-gas turbine power plant, Rankine steam cycle and ammonia-water absorption refrigeration system is introduced and analyzed. In this process, power, heat and cooling are produced. Energy and exergy analyses along with the economic factors are used to distinguish optimum operating point of the system. The developed electrochemical model of the fuel cell is validated with experimental results. Thermodynamic package and main parameters of the absorption refrigeration system are validated. The power output of the system is 500 kW. An optimization problem is defined in order to finding the optimal operating point. Decision variables are current density, temperature of the exhaust gases from the boiler, steam turbine pressure (high and medium), generator temperature and consumed cooling water. Results indicate that electrical efficiency of the combined system is 62.4% (LHV). Produced refrigeration (at -10 °C) and heat recovery are 101 kW and 22.1 kW respectively. Investment cost for the combined system (without absorption cycle) is about 2917 kW-1.

Advanced Diesel Oil Fuel Processor Development

DTIC Science & Technology

1986-06-01

water exit 29 sample quencher: gas sample line inlet 30 sample quencher: gas sample line exit 31 sample quencher: cooling water inlet 32 desulfuriser ...exit line 33, 34 desulfurimer 35 heat exchanger: process gas exit (to desulfuriser ) 38 shift reactor inlet (top) 37 shift reactor: cooling air exit

Development work for a borax internal core-catcher for a gas-cooled fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Donne, M.D.; Dorner, S.; Schumacher, G.

1978-07-01

Preliminary thermal calculations show that a corecatcher, which is able to cope with the complete meltdown of the core and blankets of a 1000-MW(electric) gas-cooled fast reactor, appears to be feasible. This core-catcher is based on borax (Na/sub 2/B/sub 4/O/sub 7/) dissolving the oxide fuel and the fission products occurring in oxide form. The borax is contained in steel boxes forming a 2.2-m-thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel (PCRV), just underneath the reactor core. After a complete meltdown accident, the fission products, in oxide form, are dispersed in the pool formedmore » by the liquid borax. The metallic fission products are contained in the steel lying below the borax pool and in contact with the water-cooled PCRV liner. The volumetric power density of the molten core is conveniently reduced as it is dissolved in the borax, and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system.« less

Experimental and theoretical study of horizontal tube bundle for passive condensation heat transfer

NASA Astrophysics Data System (ADS)

Song, Yong Jae

The research in this thesis supports the design of a horizontal tube bundle condenser for passive heat removal system in nuclear reactors. From nuclear power plant containment, condensation of steam from a steam/noncondensable gas occurs on the primary side and boiling occurs on the secondary side; thus, heat exchanger modeling is a challenge. For the purpose of this experimental study, a six-tube bundle is used, where the outer diameter, inner diameter, and length of each stainless steel tube measures 38.10mm (1.5 inches), 31.75mm (1.25 inches) and 3.96m (156 inches), respectively. The pitch to diameter ratio was determined based on information gathered from literature surveys, and the dimensions were determined from calculations and experimental data. The objective of the calculations, correlations, and experimental data was to obtain complete condensation within the tube bundle. Experimental conditions for the tests in this thesis work were determined from Design Basis Accident (DBA). The applications are for an actual Passive Containment Cooling Systems (PCCS) condenser under postulated accident conditions in future light water reactors. In this research, steady state and transient experiments were performed to investigate the effect of noncondensable gas on steam condensation inside and boiling outside a tube bundle heat exchanger. The condenser tube inlet steam mass flow rate varied from 18.0 to 48.0 g/s, the inlet pressure varied from 100 kPa to 400 kPa, and the inlet noncondensable gas mass fraction varied from 1% to 10%. The effect of the noncondensable gas was examined by comparing the tube centerline temperatures for various inlet and system conditions. As a result, it was determined that the noncondensable gas accumulated near the condensate film causing a decrease of mass and energy transfer. In addition, the effect of the inlet steam flow rate gas was investigated by comparing the tube centerline temperatures, the conclusion being that, as the inlet steam mass flow rate increased, the length required for complete condensation also increased. Comparison of tube centerline temperature profiles was also used to examine the effect of inlet pressure on the heat transfer performance. From this assessment, it was determined that as the inlet pressure increased, the length required for complete condensation decreased. The investigation of tube bundle effects was conducted by comparing the condensate flow rates. The experimental results showed that the upper tubes in the bundle had better heat transfer performance than the lower tubes. In regard to modeling of the heat exchanger in this study, for the primary side, an empirical correlation was developed herein to provide Nusselt numbers for condensation heat transfer in horizontal tubes with noncondensable gases. Nusselt numbers were correlated as: Nu = 106.31·Re m0.147·W a-0.843. The empirical model for condensation heat transfer coefficients and the secondary-side model were integrated within a Matlab program to provide an analysis tool for horizontal tube bundle condenser heat exchangers. Also on the secondary side, two phase heat transfer coefficients were modeled considering both convective boiling and nucleate boiling as: hTP = 10.03·exp(-2.28·alpha)· hCV + 0.076·exp[3.73x10-6·(Re f-1.6x105)]·hNB.

Concentrating Solar Power Projects - La Africana | Concentrating Solar

Science.gov Websites

: Posadas (Córdoba) Owner(s): Ortiz/TSK/Magtel (100%) Technology: Parabolic trough Turbine Capacity: Net -Field Outlet Temp: 393°C Solar-Field Temp Difference: 100°C Power Block Turbine Capacity (Gross): 50.0 MW Turbine Capacity (Net): 50.0 MW Output Type: Steam Rankine Cooling Method: Wet cooling Thermal

Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

NASA Astrophysics Data System (ADS)

Kim, Kwangmin; Go, Byeong-Soo; Sung, Hae-Jin; Park, Hea-chul; Kim, Seokho; Lee, Sangjin; Jin, Yoon-Su; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2014-09-01

This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

Development of a thermal scheme for a cogeneration combined-cycle unit with an SVBR-100 reactor

NASA Astrophysics Data System (ADS)

Kasilov, V. F.; Dudolin, A. A.; Krasheninnikov, S. M.

2017-02-01

At present, the prospects for development of district heating that can increase the effectiveness of nuclear power stations (NPS), cut down their payback period, and improve protection of the environment against harmful emissions are being examined in the nuclear power industry of Russia. It is noted that the efficiency of nuclear cogeneration power stations (NCPS) is drastically affected by the expenses for heat networks and heat losses during transportation of a heat carrier through them, since NPSs are usually located far away from urban area boundaries as required for radiation safety of the population. The prospects for using cogeneration power units with small or medium power reactors at NPSs, including combined-cycle units and their performance indices, are described. The developed thermal scheme of a cogeneration combined-cycle unit (CCU) with an SBVR-100 nuclear reactor (NCCU) is presented. This NCCU should use a GE 6FA gasturbine unit (GTU) and a steam-turbine unit (STU) with a two-stage district heating plant. Saturated steam from the nuclear reactor is superheated in a heat-recovery steam generator (HRSG) to 560-580°C so that a separator-superheater can be excluded from the thermal cycle of the turbine unit. In addition, supplemental fuel firing in HRSG is examined. NCCU effectiveness indices are given as a function of the ambient air temperature. Results of calculations of the thermal cycle performance under condensing operating conditions indicate that the gross electric efficiency η el NCCU gr of = 48% and N el NCCU gr = 345 MW can be achieved. This efficiency is at maximum for NCCU with an SVBR-100 reactor. The conclusion is made that the cost of NCCU installed kW should be estimated, and the issue associated with NCCUs siting with reference to urban area boundaries must be solved.

Assessment of the Use of Nitrogen Trifluoride for Purifying Coolant and Heat Transfer Salts in the Fluoride Salt-Cooled High-Temperature Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scheele, Randall D.; Casella, Andrew M.

2010-09-28

This report provides an assessment of the use of nitrogen trifluoride for removing oxide and water-caused contaminants in the fluoride salts that will be used as coolants in a molten salt cooled reactor.

NEUTRONIC REACTOR

DOEpatents

Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

1959-10-27

BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fuller, L.C.

The ORCENT-II digital computer program will perform calculations at valves-wide-open design conditions, maximum guaranteed rating conditions, and an approximation of part-load conditions for steam turbine cycles supplied with throttle steam characteristic of contemporary light-water reactors. Turbine performance calculations are based on a method published by the General Electric Company. Output includes all information normally shown on a turbine-cycle heat balance diagram. The program is written in FORTRAN IV for the IBM System 360 digital computers at the Oak Ridge National Laboratory.

Description and cost analysis of a deluge dry/wet cooling system.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wiles, L.E.; Bamberger, J.A.; Braun, D.J.

1978-06-01

The use of combined dry/wet cooling systems for large base-load power plants offers the potential for significant water savings as compared to evaporatively cooled power plants and significant cost savings in comparison to dry cooled power plants. The results of a detailed engineering and cost study of one type of dry/wet cooling system are described. In the ''deluge'' dry/wet cooling method, a finned-tube heat exchanger is designed to operate in the dry mode up to a given ambient temperature. To avoid the degradation of performance for higher ambient temperatures, water (the delugeate) is distributed over a portion of the heatmore » exchanger surface to enhance the cooling process by evaporation. The deluge system used in this study is termed the HOETERV system. The HOETERV deluge system uses a horizontal-tube, vertical-plate-finned heat exchanger. The delugeate is distributed at the top of the heat exchanger and is allowed to fall by gravity in a thin film on the face of the plate fin. Ammonia is used as the indirect heat transfer medium between the turbine exhaust steam and the ambient air. Steam is condensed by boiling ammonia in a condenser/reboiler. The ammonia is condensed in the heat exchanger by inducing airflow over the plate fins. Various design parameters of the cooling system have been studied to evaluate their impact on the optimum cooling system design and the power-plant/utility-system interface. Annual water availability was the most significant design parameter. Others included site meteorology, heat exchanger configuration and air flow, number and size of towers, fan system design, and turbine operation. It was concluded from this study that the HOETERV deluge system of dry/wet cooling, using ammonia as an intermediate heat transfer medium, offers the potential for significant cost savings compared with all-dry cooling, while achieving substantially reduced water consumption as compared to an evaporatively cooled power plant. (LCL)« less

Condensation heat transfer and flow friction in silicon microchannels

NASA Astrophysics Data System (ADS)

Wu, Huiying; Wu, Xinyu; Qu, Jian; Yu, Mengmeng

2008-11-01

An experimental investigation was performed on heat transfer and flow friction characteristics during steam condensation flow in silicon microchannels. Three sets of trapezoidal silicon microchannels, with hydraulic diameters of 77.5 µm, 93.0 µm and 128.5 µm respectively, were tested under different flow and cooling conditions. It was found that both the condensation heat transfer Nusselt number (Nu) and the condensation two-phase frictional multiplier (phi2Lo) were dependent on the steam Reynolds number (Rev), condensation number (Co) and dimensionless hydraulic diameter (Dh/L). With the increase in the steam Reynolds number, condensation number and dimensionless hydraulic diameter, the condensation Nusselt number increased. However, different variations were observed for the condensation two-phase frictional multiplier. With the increase in the steam Reynolds number and dimensionless hydraulic diameter, the condensation two-phase frictional multiplier decreased, while with the increase in the condensation number, the condensation two-phase frictional multiplier increased. Based on the experimental results, dimensionless correlations for condensation heat transfer and flow friction in silicon microchannels were proposed for the first time. These correlations can be used to determine the condensation heat transfer coefficient and pressure drop in silicon microchannels if the steam mass flow rate, cooling rate and geometric parameters are fixed. It was also found that the condensation heat transfer and flow friction have relations to the injection flow (a transition flow pattern from the annular flow to the slug/bubbly flow), and with injection flow moving toward the outlet, both the condensation heat transfer coefficient and the condensation two-phase frictional multiplier increased.

Thermal-Hydraulic Design of a Fluoride High-Temperature Demonstration Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 smallmore » 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.« less

Emergency core cooling system

DOEpatents

Schenewerk, William E.; Glasgow, Lyle E.

1983-01-01

A liquid metal cooled fast breeder reactor provided with an emergency core cooling system includes a reactor vessel which contains a reactor core comprising an array of fuel assemblies and a plurality of blanket assemblies. The reactor core is immersed in a pool of liquid metal coolant. The reactor also includes a primary coolant system comprising a pump and conduits for circulating liquid metal coolant to the reactor core and through the fuel and blanket assemblies of the core. A converging-diverging venturi nozzle with an intermediate throat section is provided in between the assemblies and the pump. The intermediate throat section of the nozzle is provided with at least one opening which is in fluid communication with the pool of liquid sodium. In normal operation, coolant flows from the pump through the nozzle to the assemblies with very little fluid flowing through the opening in the throat. However, when the pump is not running, residual heat in the core causes fluid from the pool to flow through the opening in the throat of the nozzle and outwardly through the nozzle to the assemblies, thus providing a means of removing decay heat.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

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

SIMULTANEOUS DIFFERENTIAL EQUATION COMPUTER

DOEpatents

Collier, D.M.; Meeks, L.A.; Palmer, J.P.

1960-05-10

A description is given for an electronic simulator for a system of simultaneous differential equations, including nonlinear equations. As a specific example, a homogeneous nuclear reactor system including a reactor fluid, heat exchanger, and a steam boiler may be simulated, with the nonlinearity resulting from a consideration of temperature effects taken into account. The simulator includes three operational amplifiers, a multiplier, appropriate potential sources, and interconnecting R-C networks.

Core cooling under accident conditions at the high-flux beam reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tichler, P.; Cheng, L.; Fauske, H.

The High-Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is cooled and moderated by heavy water and contains {sup 235}U in the form of narrow-channel, parallel-plate-type fuel elements. During normal operation, the flow direction is downward through the core. This flow direction is maintained at a reduced flow rate during routine shutdown and on loss of commercial power by means of redundant pumps and power supplies. However, in certain accident scenarios, e.g. loss-of-coolant accidents (LOCAs), all forced-flow cooling is lost. Although there was experimental evidence during the reactor design period (1958-1963) that the heat removal capacity in the fullymore » developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. Accordingly, a test program was initiated using an electrically heated section to simulate the fuel channel and a cooling loop to simulate the balance of the primary cooling system.« less

Case Study for the ARRA-Funded Ground Source Heat Pump Demonstration at Ball State University

DOE Office of Scientific and Technical Information (OSTI.GOV)

Im, Piljae; Liu, Xiaobing; Henderson, Jr., Hugh

With funding provided by the American Recovery and Reinvestment Act (ARRA), 26 ground-source heat pump (GSHP) projects were competitively selected in 2009 to demonstrate the benefits of GSHP systems and innovative technologies for cost reduction and/or performance improvement. One of the selected demonstration projects is a district central GSHP system installed at Ball State University (BSU) in Muncie, IN. Prior to implementing the district GSHP system, 47 major buildings in BSU were served by a central steam plant with four coal-fired and three natural-gas-fired steam boilers. Cooling was provided by five water-cooled centrifugal chillers at the District Energy Station Southmore » (DESS). The new district GSHP system replaced the existing coal-fired steam boilers and conventional water-cooled chillers. It uses ground-coupled heat recovery (HR) chillers to meet the simultaneous heating and cooling demands of the campus. The actual performance of the GSHP system was analyzed based on available measured data from August 2015 through July 2016, construction drawings, maintenance records, personal communications, and construction costs. Since Phase 1 was funded in part by the ARRA grant, it is the focus of this case study. The annual energy consumption of the GSHP system was calculated based on the available measured data and other related information. It was compared with the performance of a baseline scenario— a conventional water-cooled chiller and natural-gas-fired boiler system, both of which meet the minimum energy efficiencies allowed by the American Society of Heating, Refrigerating and Air-Conditioning Engineers (ASHRAE 90.1-2013). The comparison was made to determine source energy savings, energy cost savings, and CO2 emission reductions achieved by the GSHP system. A cost analysis was performed to evaluate the simple payback of the GSHP system. The following sections summarize the results of the analysis, the lessons learned, and recommendations for improvement in the operation of this district GSHP system.« less

A 100-kWt NaK-Cooled Space Reactor Concept for an Early-Flight Mission

NASA Astrophysics Data System (ADS)

Poston, David I.

2003-01-01

A stainless-steel (SS) sodium-potassium (NaK) cooled reactor could potentially be the first step in utilizing fission technology in space. The sum of all system-level experience for liquid-metal-cooled space reactors has been with NaK, including the SNAP-10a, the only reactor ever launched by the US. This paper describes a 100-kWt NaK reactor, the NaK-100, which is designed to be developed with minimal technical risk. In additional to NaK technology heritage, the NaK-100 uses a proven fuel-form (SS/UO2) and is designed for simplified system integration and testing. The pins are placed within a solid SS prism, and the NaK flows in an annulus between the pins and the prism. The nuclear and thermal-hydraulic performance of the NaK-100 is presented, as well as the major differences between the NaK-100 and SNAP-10a.

PBF Cooling Tower. View from highbay roof of Reactor Building ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Cooling Tower. View from high-bay roof of Reactor Building (PER-620). Camera faces northwest. East louvered face has been installed. Inlet pipes protrude from fan deck. Two redwood vents under construction at top. Note piping, control, and power lines at sub-grade level in trench leading to Reactor Building. Photographer: Kirsh. Date: June 6, 1969. INEEL negative no. 69-3466 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Multi-Megawatt Power System Trade Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Longhurst, Glen Reed; Schnitzler, Bruce Gordon; Parks, Benjamin Travis

2001-11-01

As part of a larger task, the Idaho National Engineering and Environmental Laboratory (INEEL) was tasked to perform a trade study comparing liquid-metal cooled reactors having Rankine power conversion systems with gas-cooled reactors having Brayton power conversion systems. This report summarizes the approach, the methodology, and the results of that trade study. Findings suggest that either approach has the possibility to approach the target specific mass of 3-5 kg/kWe for the power system, though it appears either will require improvements to achieve that. Higher reactor temperatures have the most potential for reducing the specific mass of gas-cooled reactors but domore » not necessarily have a similar effect for liquid-cooled Rankine systems. Fuels development will be the key to higher reactor operating temperatures. Higher temperature turbines will be important for Brayton systems. Both replacing lithium coolant in the primary circuit with gallium and replacing potassium with sodium in the power loop for liquid systems increase system specific mass. Changing the feed pump turbine to an electric motor in Rankine systems has little effect. Key technologies in reducing specific mass are high reactor and radiator operating temperatures, low radiator areal density, and low turbine/generator system masses. Turbine/generator mass tends to dominate overall power system mass for Rankine systems. Radiator mass was dominant for Brayton systems.« less

AIR COOLED NEUTRONIC REACTOR

DOEpatents

Fermi, E.; Szilard, L.

1958-05-27

A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

Solar heated oil shale pyrolysis process

NASA Technical Reports Server (NTRS)

Qader, S. A. (Inventor)

1985-01-01

An improved system for recovery of a liquid hydrocarbon fuel from oil shale is presented. The oil shale pyrolysis system is composed of a retort reactor for receiving a bed of oil shale particules which are heated to pyrolyis temperature by means of a recycled solar heated gas stream. The gas stream is separated from the recovered shale oil and a portion of the gas stream is rapidly heated to pyrolysis temperature by passing it through an efficient solar heater. Steam, oxygen, air or other oxidizing gases can be injected into the recycle gas before or after the recycle gas is heated to pyrolysis temperature and thus raise the temperature before it enters the retort reactor. The use of solar thermal heat to preheat the recycle gas and optionally the steam before introducing it into the bed of shale, increases the yield of shale oil.

Solar heated fluidized bed gasification system

NASA Technical Reports Server (NTRS)

Qader, S. A. (Inventor)

1981-01-01

A solar-powered fluidized bed gasification system for gasifying carbonaceous material is presented. The system includes a solar gasifier which is heated by fluidizing gas and steam. Energy to heat the gas and steam is supplied by a high heat capacity refractory honeycomb which surrounds the fluid bed reactor zone. The high heat capacity refractory honeycomb is heated by solar energy focused on the honeycomb by solar concentrator through solar window. The fluid bed reaction zone is also heated directly and uniformly by thermal contact of the high heat capacity ceramic honeycomb with the walls of the fluidized bed reactor. Provisions are also made for recovering and recycling catalysts used in the gasification process. Back-up furnace is provided for start-up procedures and for supplying heat to the fluid bed reaction zone when adequate supplies of solar energy are not available.

Study on steam pressure characteristics in various types of nozzles

NASA Astrophysics Data System (ADS)

Firman; Anshar, Muhammad

2018-03-01

Steam Jet Refrigeration (SJR) is one of the most widely applied technologies in the industry. The SJR system was utilizes residual steam from the steam generator and then flowed through the nozzle to a tank that was containing liquid. The nozzle converts the pressure energy into kinetic energy. Thus, it can evaporate the liquid briefly and release it to the condenser. The chilled water, was produced from the condenser, can be used to cool the product through a heat transfer process. This research aims to study the characteristics of vapor pressure in different types of nozzles using a simulation. The Simulation was performed using ANSYS FLUENT software for nozzle types such as convergent, convrgent-parallel, and convergent-divergent. The results of this study was presented the visualization of pressure in nozzles and was been validated with experiment data.