Effect of Silver or Copper Nanoparticles-Dispersed Silane Coatings on Biofilm Formation in Cooling Water Systems

PubMed Central

Ogawa, Akiko; Kanematsu, Hideyuki; Sano, Katsuhiko; Sakai, Yoshiyuki; Ishida, Kunimitsu; Beech, Iwona B.; Suzuki, Osamu; Tanaka, Toshihiro

2016-01-01

Biofouling often occurs in cooling water systems, resulting in the reduction of heat exchange efficiency and corrosion of the cooling pipes, which raises the running costs. Therefore, controlling biofouling is very important. To regulate biofouling, we focus on the formation of biofilm, which is the early step of biofouling. In this study, we investigated whether silver or copper nanoparticles-dispersed silane coatings inhibited biofilm formation in cooling systems. We developed a closed laboratory biofilm reactor as a model of a cooling pipe and used seawater as a model for cooling water. Silver or copper nanoparticles-dispersed silane coating (Ag coating and Cu coating) coupons were soaked in seawater, and the seawater was circulated in the laboratory biofilm reactor for several days to create biofilms. Three-dimensional images of the surface showed that sea-island-like structures were formed on silane coatings and low concentration Cu coating, whereas nothing was formed on high concentration Cu coatings and low concentration Ag coating. The sea-island-like structures were analyzed by Raman spectroscopy to estimate the components of the biofilm. We found that both the Cu coating and Ag coating were effective methods to inhibit biofilm formation in cooling pipes. PMID:28773758

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.

Space Nuclear Reactor Engineering

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

Poston, David Irvin

We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

Promethus Hot Leg Piping Concept

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

AM Girbik; PA Dilorenzo

2006-01-24

The Naval Reactors Prime Contractor Team (NRPCT) recommended the development of a gas cooled reactor directly coupled to a Brayton energy conversion system as the Space Nuclear Power Plant (SNPP) for NASA's Project Prometheus. The section of piping between the reactor outlet and turbine inlet, designated as the hot leg piping, required unique design features to allow the use of a nickel superalloy rather than a refractory metal as the pressure boundary. The NRPCT evaluated a variety of hot leg piping concepts for performance relative to SNPP system parameters, manufacturability, material considerations, and comparison to past high temperature gas reactormore » (HTGR) practice. Manufacturability challenges and the impact of pressure drop and turbine entrance temperature reduction on cycle efficiency were discriminators between the piping concepts. This paper summarizes the NRPCT hot leg piping evaluation, presents the concept recommended, and summarizes developmental issues for the recommended concept.« less

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.

Application of Molten Salt Reactor Technology to Nuclear Electric Propulsion Mission

NASA Technical Reports Server (NTRS)

Patton, Bruce; Sorensen, Kirk; Rodgers, Stephen L. (Technical Monitor)

2002-01-01

Nuclear electric propulsion (NEP) and planetary surface power missions require reactors that are lightweight, operationally robust, and scalable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional gas cooled, liquid metal, and heat pipe space reactors.

Thermionic nuclear reactor with internal heat distribution and multiple duct cooling

DOEpatents

Fisher, C.R.; Perry, L.W. Jr.

1975-11-01

A Thermionic Nuclear Reactor is described having multiple ribbon-like coolant ducts passing through the core, intertwined among the thermionic fuel elements to provide independent cooling paths. Heat pipes are disposed in the core between and adjacent to the thermionic fuel elements and the ribbon ducting, for the purpose of more uniformly distributing the heat of fission among the thermionic fuel elements and the ducts.

ETR, TRA642. EASTWEST SECTION, LOOKING NORTH. PATH OF COOLING WATER ...

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

ETR, TRA-642. EAST-WEST SECTION, LOOKING NORTH. PATH OF COOLING WATER PIPE TUNNEL. WORKING AND STORAGE CANAL. SUB-PILE ROOM. CONTROL ROD ACCESS ROOM. FLOOR NAMES. (THIS WAS A CONCEPT DRAWING.) KAISER ETR-5528-MTR-642-A-5, 11/1955. INL INDEX NO. 532-0642-00-486-100913. REV. 0. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

High Temperature Water Heat Pipes Radiator for a Brayton Space Reactor Power System

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.; Tournier, Jean-Michel

2006-01-01

A high temperature water heat pipes radiator design is developed for a space power system with a sectored gas-cooled reactor and three Closed Brayton Cycle (CBC) engines, for avoidance of single point failures in reactor cooling and energy conversion and rejection. The CBC engines operate at turbine inlet and exit temperatures of 1144 K and 952 K. They have a net efficiency of 19.4% and each provides 30.5 kWe of net electrical power to the load. A He-Xe gas mixture serves as the turbine working fluid and cools the reactor core, entering at 904 K and exiting at 1149 K. Each CBC loop is coupled to a reactor sector, which is neutronically and thermally coupled, but hydraulically decoupled to the other two sectors, and to a NaK-78 secondary loop with two water heat pipes radiator panels. The segmented panels each consist of a forward fixed segment and two rear deployable segments, operating hydraulically in parallel. The deployed radiator has an effective surface area of 203 m2, and when the rear segments are folded, the stowed power system fits in the launch bay of the DELTA-IV Heavy launch vehicle. For enhanced reliability, the water heat pipes operate below 50% of their wicking limit; the sonic limit is not a concern because of the water, high vapor pressure at the temperatures of interest (384 - 491 K). The rejected power by the radiator peaks when the ratio of the lengths of evaporator sections of the longest and shortest heat pipes is the same as that of the major and minor widths of the segments. The shortest and hottest heat pipes in the rear segments operate at 491 K and 2.24 MPa, and each rejects 154 W. The longest heat pipes operate cooler (427 K and 0.52 MPa) and because they are 69% longer, reject more power (200 W each). The longest and hottest heat pipes in the forward segments reject the largest power (320 W each) while operating at ~ 46% of capillary limit. The vapor temperature and pressure in these heat pipes are 485 K and 1.97 MPa. By contrast, the shortest water heat pipes in the forward segments operate much cooler (427 K and 0.52 MPa), and reject a much lower power of 45 W each. The radiator with six fixed and 12 rear deployable segments rejects a total of 324 kWth, weights 994 kg and has an average specific power of 326 Wth/kg and a specific mass of 5.88 kg/m2.

Dynamic System Simulation of the KRUSTY Experiment

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

Klein, Steven Karl; Kimpland, Robert Herbert

2016-05-09

The proposed KRUSTY experiment is a demonstration of a reactor operating at power. The planned experimental configuration includes a highly enriched uranium (HEU) reflected core, cooled by multiple heat pipes leading to Stirling engines for primary heat rejection. Operating power is expected to be approximately four (4) to five (5) kilowatts with a core temperature above 1,000 K. No data is available on any historical reactor employing HEU metal that operated over the temperature range required for the KRUSTY experiment. Further, no reactor has operated with heat pipes as the primary cooling mechanism. Historic power reactors have employed either naturalmore » or forced convection so data on their operation is not directly applicable to the KRUSTY experiment. The primary purpose of the system model once developed and refined by data from these component experiments, will be used to plan the KRUSTY experiment. This planning will include expected behavior of the reactor from start-up, through various transient conditions where cooling begins to become present and effective, and finally establishment of steady-state. In addition, the model can provide indicators of anticipated off-normal events and appropriate operator response to those conditions. This information can be used to develop specific experiment operating procedures and aids to guide the operators in conduct of the experiment.« less

Proposed Design and Operation of a Heat Pipe Reactor using the Sandia National Laboratories Annular Core Test Facility and Existing UZrH Fuel Pins

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

Wright, Steven A.; Lipinski, Ronald J.; Pandya, Tara

2005-02-06

Heat Pipe Reactors (HPR) for space power conversion systems offer a number of advantages not easily provided by other systems. They require no pumping, their design easily deals with freezing and thawing of the liquid metal, and they can provide substantial levels of redundancy. Nevertheless, no reactor has ever been operated and cooled with heat pipes, and the startup and other operational characteristics of these systems remain largely unknown. Signification deviations from normal reactor heat removal mechanisms exist, because the heat pipes have fundamental heat removal limits due to sonic flow issues at low temperatures. This paper proposes an earlymore » prototypic test of a Heat Pipe Reactor (using existing 20% enriched nuclear fuel pins) to determine the operational characteristics of the HPR. The proposed design is similar in design to the HOMER and SAFE-300 HPR designs (Elliot, Lipinski, and Poston, 2003; Houts, et. al, 2003). However, this reactor uses existing UZrH fuel pins that are coupled to potassium heat pipes modules. The prototype reactor would be located in the Sandia Annular Core Research Reactor Facility where the fuel pins currently reside. The proposed reactor would use the heat pipes to transport the heat from the UZrH fuel pins to a water pool above the core, and the heat transport to the water pool would be controlled by adjusting the pressure and gas type within a small annulus around each heat pipe. The reactor would operate as a self-critical assembly at power levels up to 200 kWth. Because the nuclear heated HPR test uses existing fuel and because it would be performed in an existing facility with the appropriate safety authorization basis, the test could be performed rapidly and inexpensively. This approach makes it possible to validate the operation of a HPR and also measure the feedback mechanisms for a typical HPR design. A test of this nature would be the world's first operating Heat Pipe Reactor. This reactor is therefore called 'HPR-1'.« less

Transient Approximation of SAFE-100 Heat Pipe Operation

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Reid, Robert S.

2005-01-01

Engineers at Los Alamos National Laboratory (LANL) have designed several heat pipe cooled reactor concepts, ranging in power from 15 kWt to 800 kWt, for both surface power systems and nuclear electric propulsion systems. The Safe, Affordable Fission Engine (SAFE) is now being developed in a collaborative effort between LANL and NASA Marshall Space Flight Center (NASA/MSFC). NASA is responsible for fabrication and testing of non-nuclear, electrically heated modules in the Early Flight Fission Test Facility (EFF-TF) at MSFC. In-core heat pipes must be properly thawed as the reactor power starts. Computational models have been developed to assess the expected operation of a specific heat pipe design during start-up, steady state operation, and shutdown. While computationally intensive codes provide complete, detailed analyses of heat pipe thaw, a relatively simple. concise routine can also be applied to approximate the response of a heat pipe to changes in the evaporator heat transfer rate during start-up and power transients (e.g., modification of reactor power level) with reasonably accurate results. This paper describes a simplified model of heat pipe start-up that extends previous work and compares the results to experimental measurements for a SAFE-100 type heat pipe design.

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.

Exploring the engineering limit of heat flux of a W/RAFM divertor target for fusion reactors

NASA Astrophysics Data System (ADS)

Mao, X.; Fursdon, M.; Chang, X. B.; Zhang, J. W.; Liu, P.; Ellwood, G.; Qian, X. Y.; Qin, S. J.; Peng, X. B.; Barrett, T. R.; Liu, P.

2018-06-01

The design and development of a fusion reactor divertor plasma facing component (PFC) is one of the many challenging issues on the road to commercial use of fusion energy. The divertor PFC is expected to exhaust steady state heat loads in the region of 10 MW m‑2 while keeping temperatures and thermo-mechanical stresses in its structure within the allowable limits. For ITER (International Thermo-Nuclear Experimental Reactor) a water cooled W/CuCrZr divertor PFC concept has been developed. However, this concept is not necessarily assured for use in future fusion reactors mainly because the neutron radiation dose would be at least an order magnitude higher, resulting in limited thermo-mechanical performance and considerably more activated waste products. In the present study, a water cooled divertor PFC using reduced activation ferritic-martensitic (RAFM) steel as the heat sink pipe has been designed with pressurised water reactor-like cooling conditions (pressure of 15.5 MPa, velocity of 10–20 m s‑1 and temperature of 300 °C). The PFC is made up of a number of rectangular tungsten tiles, each with an inner circular hole (so-called monoblocks), joined onto a RAFM steel pipe with copper interlayers. The thermo-mechanical performance of the PFC has been studied in detail. The heat transfer coefficient between the RAFM pipe inner surface and the water was calculated using published correlations. Geometric parameters and water velocity were optimized with finite element (FE) thermal analysis, to achieve acceptable temperatures in the structure given the target exhaust heat load of 10 MW m‑2. Under this heat load and the optimised thermal design parameters, the structure of the PFC was further assessed by mechanical analysis. We find that under these conditions the RAFM steel pipe experiences cyclic plasticity, and fails the common linear elastic ratchetting (3 Sm) rule. Nevertheless, the designed W/RAFM divertor PFU can withstand 10 MW m‑2 heat load, albeit with a fatigue life of approximately 0.55 years based on the expected operation scenario of a prototype or test reactor. This study extends the state of knowledge of the technological limit of a divertor based on a RAFM steel pipe structure.

Small Reactor for Deep Space Exploration

ScienceCinema

none,

2018-06-06

This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

A PC-based high temperature gas reactor simulator for Indonesian conceptual HTR reactor basic training

NASA Astrophysics Data System (ADS)

Syarip; Po, L. C. C.

2018-05-01

In planning for nuclear power plant construction in Indonesia, helium cooled high temperature reactor (HTR) is favorable for not relying upon water supply that might be interrupted by earthquake. In order to train its personnel, BATAN has cooperated with Micro-Simulation Technology of USA to develop a 200 MWt PC-based simulation model PCTRAN/HTR. It operates in Win10 environment with graphic user interface (GUI). Normal operation of startup, power maneuvering, shutdown and accidents including pipe breaks and complete loss of AC power have been conducted. A sample case of safety analysis simulation to demonstrate the inherent safety features of HTR was done for helium pipe break malfunction scenario. The analysis was done for the variation of primary coolant pipe break i.e. from 0,1% - 0,5 % and 1% - 10 % helium gas leakages, while the reactor was operated at the maximum constant power of 10 MWt. The result shows that the highest temperature of HTR fuel centerline and coolant were 1150 °C and 1296 °C respectively. With 10 kg/s of helium flow in the reactor core, the thermal power will back to the startup position after 1287 s of helium pipe break malfunction.

Development of Inspection and Repair Technology for Heat Exchanger Tubes in Fast Breeder Reactors

DTIC Science & Technology

2009-06-01

Technology for Heat Exchanger Tubes in Fast Breeder Reactors Akihiko NISHIMURA *1 , Takahisa SHOBU, Kiyoshi OKA, Toshihiko YAMAGUCHI, Yukihiro SHIMADA...fast breeder reactors (FBRs). It comprises a laser processing head combined with an eddy current testing unit. Ultrashort laser pulse ablation is used...be applied in the main- tenance of large structures such as nuclear reactors and chemical factories [1]. Internal access to a blanket cooling pipe

PBF Cooling Tower under construction. Cold water basin is five ...

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

PBF Cooling Tower under construction. Cold water basin is five feet deep. Foundation and basin walls are reinforced concrete. Camera facing west. Pipe openings through wall in front are outlets for return flow of cool water to reactor building. Photographer: John Capek. Date: September 4, 1968. INEEL negative no. 68-3473 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Multiple Restart Testing of a Stainless Steel Sodium Heat Pipe Module

NASA Astrophysics Data System (ADS)

Martin, James; Mireles, Omar; Reid, Robert

2005-02-01

A heat pipe cooled reactor is one of several candidate reactor concepts being considered for space power and propulsion systems to support future space exploration activities. Long life heat pipe modules, with concepts verified through a combination of theoretical analysis and experimental evaluations, would be necessary to establish the viability of this option. A number of stainless steel/sodium heat pipe modules have been designed and fabricated to support experimental testing of a Safe Affordable Fission Engine (SAFE) project, a 100-kWt core design pursued jointly by the Marshall Space Flight Center and the Los Alamos National Laboratory. One of the SAFE heat pipe modules was successfully subjected to over 200 restarts, examining the behavior of multiple passive freeze/thaw operations. Typical operation included a 1-hour startup to an average evaporator temperature of 1000 K followed by a 15-minute hold at temperature. Nominal maximum input power to the evaporator (measured at the power supply) during the hold period was 1.9 kW, with approximately 1.6 kW calculated as the axial power transfer to the condenser (the 300W difference was lost to environment at the evaporator surface). Between heating cycles the module was cooled to less than 325 K, returning the sodium to a frozen state in preparation for the next startup cycle.

Multiple Restart Testing of a Stainless Steel Sodium Heat Pipe Module

NASA Technical Reports Server (NTRS)

Martin, James; Mireles, Omar; Reid, Robert

2005-01-01

A heat pipe cooled reactor is one of several candidate reactor cores being considered for space power and propulsion systems to support future space exploration activities. Long life heat pipe modules. with designs verified through a combination of theoretical analysis and experimental evaluations. would be necessary to establish the viability of this option. A hardware-based program was initiated to begin experimental testing of components to verify compliance of proposed designs. To this end, a number of stainless steel/sodium heat pipe modules have been designed and fabricated to support experimental testing of a Safe Affordable Fission Engine (SAFE) project, a 100-kWt core design pursued jointly by the Marshall Space Flight Center and the Los Alamos National Laboratory. One of the SAFE heat pipe modules was successfully subjected to over 200 restarts. examining the behavior of multiple passive freeze/thaw operations. Typical operation included a 1-hour startup to an average evaporator temperature of 1000 K followed by a 15 minute hold at temperature. Nominal maximum input power during the hold period was 1.9 kW. Between heating cycles the module was cooled to less than 325 K, returning the sodium to a frozen state in preparation fop the next startup cycle.

Study of the collector/heat pipe cooled externally configured thermionic diode

NASA Technical Reports Server (NTRS)

1973-01-01

A collector/heat pipe cooled, externally configured (heated) thermionic diode module was designed for use in a laboratory test to demonstrate the applicability of this concept as the fuel element/converter module of an in-core thermionic electric power source. During the course of the program, this module evolved from a simple experimental mock-up into an advanced unit which was more reactor prototypical. Detailed analysis of all diode components led to their engineering design, fabrication, and assembly, with the exception of the collector/heat pipe. While several designs of high power annular wicked heat pipes were fabricated and tested, each exhibited unexpected performance difficulties. It was concluded that the basic cause of these problems was the formation of crud which interfered with the liquid flow in the annular passage of the evaporator region.

PROCESS WATER BUILDING, TRA605. CAMERA LOOKING EAST AND TO WEST ...

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

PROCESS WATER BUILDING, TRA-605. CAMERA LOOKING EAST AND TO WEST WALL NOW ENCLOSING FLASH EVAPORATORS. PIPES IN FOREGROUND WILL CARRY DEMINERALIZED COOLING WATER TO AND FROM THE MTR. INL NEGATIVE NO. 2937. Unknown Photographer, 7/30/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PBF Cooling Tower (PER720) and its Auxiliary Building (PER625). Camera ...

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

PBF Cooling Tower (PER-720) and its Auxiliary Building (PER-625). Camera facing west shows east facades. Center pipe carried secondary coolant water from reactor. Building to distributor basin. Date: August 2003. INEEL negative no. HD-35-10-1 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

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.

Nuclear safety considerations in the conceptual design of a fast reactor for space electric power and propulsion

NASA Technical Reports Server (NTRS)

Hsieh, T.-M.; Koenig, D. R.

1977-01-01

Some nuclear safety aspects of a 3.2 mWt heat pipe cooled fast reactor with out-of-core thermionic converters are discussed. Safety related characteristics of the design including a thin layer of B4C surrounding the core, the use of heat pipes and BeO reflector assembly, the elimination of fuel element bowing, etc., are highlighted. Potential supercriticality hazards and countermeasures are considered. Impacts of some safety guidelines of space transportation system are also briefly discussed, since the currently developing space shuttle would be used as the primary launch vehicle for the nuclear electric propulsion spacecraft.

Modeling of transient heat pipe operation

NASA Technical Reports Server (NTRS)

Colwell, Gene T.

1987-01-01

The use of heat pipes is being considered as a means of reducing the peak temperature and large thermal gradients at the leading edges of reentry vehicles and hypersonic aircraft and in nuclear reactors. In the basic cooling concept, the heat pipe covers the leading edge, a portion of the lower wing surface, and a portion of the upper wing surface. Aerodynamic heat is mainly absorbed at the leading edge and transported through the heat pipe to the upper and lower wing surface, where it is rejected by thermal radiation and convection. Basic governing equations are written to determine the startup, transient, and steady state performance of a haet pipe which has initially frozen alkali-metal as the working fluid.

Closeout Report for the Refractory Metal Accelerated Heat Pipe Life Test Activity

NASA Technical Reports Server (NTRS)

Martin, J.; Reid, R.; Stewart, E.; Hickman, R.; Mireles, O.

2013-01-01

With the selection of a gas-cooled reactor, this heat pipe accelerated life test activity was closed out and its resources redirected. The scope of this project was to establish the long-term aging effects on Mo-44.5%Re sodium heat pipes when subjected to space reactor temperature and mass fluences. To date, investigators have demonstrated heat pipe life tests of alkali metal systems up to .50,000 hours. Unfortunately, resources have not been available to examine the effect of temperature, mass fluence, or impurity level on corrosion or to conduct post-test forensic examination of heat pipes. The key objective of this effort was to establish a cost/time effective method to systematically test alkali metal heat pipes with both practical and theoretical benefits. During execution of the project, a heat pipe design was established, a majority of the laboratory test equipment systems specified, and operating and test procedures developed. Procurements for the heat pipe units and all major test components were underway at the time the stop work order was issued. An extremely important outcome was the successful fabrication of an annular wick from Mo-5%Re screen (the single, most difficult component to manufacture) using a hot isostatic pressing technique. This Technical Publication (TP) includes specifics regarding the heat pipe calorimeter water-cooling system, vendor design for the radio frequency heating system, possible alternative calorimeter designs, and progress on the vanadium equilibration technique. The methods provided in this TP and preceding project documentation would serve as a good starting point to rapidly implement an accelerated life test. Relevant test data can become available within months, not years, and destructive examination of the first life test heat pipe might begin within 6 months of test initiation. Final conclusions could be drawn in less than a quarter of the mission duration for a long-lived, fission-powered, deep space probe.

Emergency heat removal system for a nuclear reactor

DOEpatents

Dunckel, Thomas L.

1976-01-01

A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

Heat Pipe Powered Stirling Conversion for the Demonstration Using Flattop Fission (DUFF) Test

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Briggs, Maxwell H.; Sanzi, James L.; Brace, Michael H.

2013-01-01

Design concepts for small Fission Power Systems (FPS) have shown that heat pipe cooled reactors provide a passive, redundant, and lower mass option to transfer heat from the fuel to the power conversion system, as opposed to pumped loop designs typically associated with larger FPS. Although many systems have been conceptually designed and a few making it to electrically heated testing, none have been coupled to a real nuclear reactor. A demonstration test named DUFF Demonstration Using Flattop Fission, was planned by the Los Alamos National Lab (LANL) to use an existing criticality experiment named Flattop to provide the nuclear heat source. A team from the NASA Glenn Research Center designed, built, and tested a heat pipe and power conversion system to couple to Flattop with the end goal of making electrical power. This paper will focus on the design and testing performed in preparation for the DUFF test.

PBF Reactor Building (PER620). Camera faces south along west wall. ...

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

PBF Reactor Building (PER-620). Camera faces south along west wall. Gap between native lava rock and concrete basement walls is being backfilled and compacted. Wire mesh protects workers from falling rock. Note penetrations for piping that will carry secondary coolant water to Cooling Tower. Photographer: Holmes. Date: June 15, 1967. INEEL negative no. 67-3665 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Space power reactor in-core thermionic multicell evolutionary (S-prime) design

NASA Astrophysics Data System (ADS)

Determan, William R.; Van Hagan, Tom H.

1993-01-01

A 5- to 40-kWe moderated in-core thermionic space nuclear power system (TI-SNPS) concept was developed to address the TI-SNPS program requirements. The 40-kWe baseline design uses multicell Thermionic Fuel Elements (TFEs) in a zirconium hydride moderated reactor to achieve a specific mass of 18.2 We/kg and a net end-of-mission (EOM) efficiency of 8.2%. The reactor is cooled with a single NaK-78 pumped loop, which rejects the heat through a 24 m2 heat pipe space radiator.

Combination pipe-rupture mitigator and in-vessel core catcher. [LMFBR

DOEpatents

Tilbrook, R.W.; Markowski, F.J.

1982-03-09

A device is described which mitigates against the effects of a failed coolant loop in a nuclear reactor by restricting the outflow of coolant from the reactor through the failed loop and by retaining any particulated debris from a molten core which may result from coolant loss or other cause. The device reduces the reverse pressure drop through the failed loop by limiting the access of coolant in the reactor to the inlet of the failed loop. The device also spreads any particulated core debris over a large area to promote cooling.

Combination pipe rupture mitigator and in-vessel core catcher

DOEpatents

Tilbrook, Roger W.; Markowski, Franz J.

1983-01-01

A device which mitigates against the effects of a failed coolant loop in a nuclear reactor by restricting the outflow of coolant from the reactor through the failed loop and by retaining any particulated debris from a molten core which may result from coolant loss or other cause. The device reduces the reverse pressure drop through the failed loop by limiting the access of coolant in the reactor to the inlet of the failed loop. The device also spreads any particulated core debris over a large area to promote cooling.

Modeling of transient heat pipe operation

NASA Technical Reports Server (NTRS)

Colwell, G. T.; Hartley, J. G.

1986-01-01

Mathematical models and associated solution procedures which can be used to design heat pipe cooled structures for use on hypersonic vehicles are being developed. The models should also have the capability to predict off-design performance for a variety of operating conditions. It is expected that the resulting models can be used to predict startup behavior of liquid metal heat pipes to be used in reentry vehicles, hypersonic aircraft, and space nuclear reactors. Work to date related to numerical solutions of governing differential equations for the outer shell and the combination capillary structure and working fluid is summarized. Finite element numerical equations using both implicit, explicit, and combination methods were examined.

APPARATUS FOR DETECTING AND LOCATING PRESENCE OF FLUIDS

DOEpatents

Williamson, R.R.

1958-09-16

A system is described fur detecting water leaks in water-cooled neutronic reactors by utilizing an electrical hygrometer having a resistance element variable with the moisture content. The graphite blocks, forming the moderator in many types of reactors, coniain ducts in which helium gas is circulated. When a leak occurs in a coolant tube, the water will seep through the graphite until it oozes into one of the helium ducts, where it will be swept along with the helium into a system of pipes that connect each of the helium ducts. By inserting an electric hygrometer in each of these pipes and connecting it to an alarm system, the moisture content of the helium will cause a change in the electrical resistance of the hygrometer which will initiate a signal alarm indicating the presence and position of the leaky water tube in the reactor.

Application of Simulated Reactivity Feedback in Nonnuclear Testing of a Direct-Drive Gas-Cooled Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Webster, K. L.

2007-01-01

Nonnuclear testing can be a valuable tool in the development of an in-space nuclear power or propulsion system. In a nonnuclear test facility, electric heaters are used to simulate heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response and response characteristics, and assess potential design improvements with a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE 100a heat pipe cooled, electrically heated reactor and heat exchanger hardware. This Technical Memorandum discusses the status of the planned dynamic test methodology for implementation in the direct-drive gas-cooled reactor testing and assesses the additional instrumentation needed to implement high-fidelity dynamic testing.

21. View from the work area of the front face ...

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

21. View from the work area of the front face of the pile in the 105 building, in this case at the F Reactor in February 1945. The 2,004 pigtails and process tube nozzles are neatly aligned in rows and columns across the face of the pile. The cooling water risers stand at the left and right of the pile and the distribution crossheaders run across its face. The pipes running vertically at the bottom of the pile carry cooling water to the thermal shield. The low railing along the floor in front of the face prevented workers from accidentally falling into the charging elevator pit. D-8326 - B Reactor, Richland, Benton County, WA

Materials technology for an advanced space power nuclear reactor concept: Program summary

NASA Technical Reports Server (NTRS)

Gluyas, R. E.; Watson, G. K.

1975-01-01

The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

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

Middleton, Bobby; Pasch, James Jay; Kruizenga, Alan Michael

2016-01-01

This report outlines the thermodynamics of a supercritical carbon dioxide (sCO 2) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO 2 code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related tomore » both Helium and to sCO 2 , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO 2> Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO 2 Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation« less

Reactor design and integration into a nuclear electric spacecraft

NASA Technical Reports Server (NTRS)

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

1978-01-01

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

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.

Maintenance method and its critical issues for a fast-ignition laser fusion reactor based on a dry wall chamber

NASA Astrophysics Data System (ADS)

Someya, Y.; Matsumoto, T.; Okano, K.; Asaoka, Y.; Hiwatari, R.; Goto, T.; Ogawa, Y.

2008-05-01

The neutronics analysis has been carried out for feasibility study of the FALCON-D concept by Monte Carlo N-paticle transport code (MCNP), in order to inspect the cooling performance of in-vessel and ex-vessel components, and a connection pipe between Vacuum Vessel and reactor room. The nuclear heating rate in the Vacuum Vessel was at the same level as that of NBI duct of the ITER. The temperature of the connection pipe was found to be 345·, ·which was smaller than the melting point of structure materials (F82H). Moreover, the radiation damage of the final optics was also investigated. We propose a sliding changer concept for replacement. This method could be adapted for the replacement of one FPY cycle in the final optics system.

Peach Bottom Atomic Power Station recirc pipe dose rates with zinc injection and condenser replacement

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

DiCello, D.C.; Odell, A.D.; Jackson, T.J.

1995-03-01

Peach Bottom Atomic Power Station (PBAPS) is located near the town of Delta, Pennsylvania, on the west bank of the Susquehanna River. It is situated approximately 20 miles south of Lancaster, Pennsylvania. The site contains two boiling water reactors of General Electric design and each rated at 3,293 megawatts thermal. The units are BWR 4s and went commercial in 1977. There is also a decommissioned high temperature gas-cooled reactor on site, Unit 1. PBAPS Unit 2 recirc pipe was replaced in 1985 and Unit 3 recirc pipes replaced in 1988 with 326 NGSS. The Unit 2 replacement pipe was electropolished,more » and the Unit 3 pipe was electropolished and passivated. The Unit 2 brass condenser was replaced with a Titanium condenser in the first quarter of 1991, and the Unit 3 condenser was replaced in the fourth quarter of 1991. The admiralty brass condensers were the source of natural zinc in both units. Zinc injection was initiated in Unit 2 in May 1991, and in Unit 3 in May 1992. Contact dose rate measurements were made in standard locations on the 28-inch recirc suction and discharge lines to determine the effectiveness of zinc injection and to monitor radiation build-up in the pipe. Additionally, HPGe gamma scans were performed to determine the isotopic composition of the oxide layer inside the pipe. In particular, the specific ({mu}Ci/cm{sup 2}) of Co-60 and Zn-65 were analyzed.« less

Reactor pressure vessel head vents and methods of using the same

DOEpatents

Gels, John L; Keck, David J; Deaver, Gerald A

2014-10-28

Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

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

Cleaning residual NaK in the fast flux test facility fuel storage cooling system

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

Burke, T.M.; Church, W.R.; Hodgson, K.M.

2008-01-15

The Fast Flux Test Facility (FFTF), located on the U.S. Department of Energy's Hanford Reservation, is a liquid metal-cooled test reactor. The FFTF was constructed to support the U.S. Liquid Metal Fast Breeder Reactor Program. The bulk of the alkali metal (sodium and NaK) has been drained and will be stored onsite prior to final disposition. Residual NaK needed to be removed from the pipes, pumps, heat exchangers, tanks, and vessels in the Fuel Storage Facility (FSF) cooling system. The cooling system was drained in 2004 leaving residual NaK in the pipes and equipment. The estimated residual NaK volume wasmore » 76 liters in the storage tank, 1.9 liters in the expansion tank, and 19-39 liters in the heat transfer loop. The residual NaK volume in the remainder of the system was expected to be very small, consisting of films, droplets, and very small pools. The NaK in the FSF Cooling System was not radiologically contaminated. The portions of the cooling system to be cleaned were divided into four groups: 1. The storage tank, filter, pump, and associated piping; 2. The heat exchanger, expansion tank, and associated piping; 3. Argon supply piping; 4. In-vessel heat transfer loop. The cleaning was contracted to Creative Engineers, Inc. (CEI) and they used their superheated steam process to clean the cooling system. It has been concluded that during the modification activities (prior to CEI coming onsite) to prepare the NaK Cooling System for cleaning, tank T-914 was pressurized relative to the In-Vessel NaK Cooler and NaK was pushed from the tank back into the Cooler and that on November 6, 2005, when the gas purge through the In-Vessel NaK Cooler was increased from 141.6 slm to 283.2 slm, NaK was forced from the In-Vessel NaK Cooler and it contacted water in the vent line and/or scrubber. The gases from the reaction then traveled back through the vent line coating the internal surface of the vent line with NaK and NaK reaction products. The hot gases also exited the scrubber through the stack and due to the temperature of the gas, the hydrogen auto ignited when it mixed with the oxygen in the air. There was no damage to equipment, no injuries, and no significant release of hazardous material. Even though the FSF Cooling System is the only system at FFTF that contains residual NaK, there are lessons to be learned from this event that can be applied to future residual sodium removal activities. The lessons learned are: - Before cleaning equipment containing residual alkali metal the volume of alkali metal in the equipment should be minimized to the extent practical. As much as possible, reconfirm the amount and location of the alkali metal immediately prior to cleaning, especially if additional evolutions have been performed or significant time has passed. This is especially true for small diameter pipe (<20.3 centimeters diameter) that is being cleaned in place since gas flow is more likely to move the alkali metal. Potential confirmation methods could include visual inspection (difficult in all-metal systems), nondestructive examination (e.g., ultrasonic measurements) and repeating previous evolutions used to drain the system. Also, expect to find alkali metal in places it would not reasonably be expected to be. - Staff with an intimate knowledge of the plant equipment and the bulk alkali metal draining activities is critical to being able to confirm the amount and locations of the alkali metal residuals and to safely clean the residuals. - Minimize the potential for movement of alkali metal during cleaning or limit the distance and locations into which alkali metal can move. - Recognize that when working with alkali metal reactions, occasional pops and bangs are to be anticipated. - Pre-plan emergency responses to unplanned events to assure responses planned for an operating reactor are appropriate for the deactivation phase.« less

SPERTI/PBF. Contextual aerial view after PBF had begun operating, but ...

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

SPERT-I/PBF. Contextual aerial view after PBF had begun operating, but prior to expansion of southwest corner of Reactor Building (PER-620). Camera facing northeast. Reactor Building in center of view. Cooling Tower (PER-720) to its left. Warehouse (PER-625) at lower left was built in 1966. SPERT-I Reactor Building (PER-605) and Instrument Cell Building (PER-604) at right of view. Buried cables and piping proceed from PBF toward lower edge of view to Control Building further south and out of view. Photographer: Farmer. Date: March 26, 1976. INEEL negative no. 76-1344 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Optimize out-of-core thermionic energy conversion for nuclear electric propulsion

NASA Technical Reports Server (NTRS)

Morris, J. F.

1977-01-01

Current designs for out of core thermionic energy conversion (TEC) to power nuclear electric propulsion (NEP) were evaluated. Approaches to improve out of core TEC are emphasized and probabilities for success are indicated. TEC gains are available with higher emitter temperatures and greater power densities. Good potentialities for accommodating external high temperature, high power density TEC with heat pipe cooled reactors exist.

RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

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

Hongbin Zhang; Haihua Zhao; Cliff Davis

2008-06-01

An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heatmore » from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet temperature. The peak cladding, hot pool, cold pool and reactor inlet temperatures were calculated during LOFC. The results indicate that there are two phases during LOFC transient – the initial thermal equilibration phase and the long term decay heat removal phase. The initial thermal equilibration phase occurs over a few hundred seconds, as the system adjusts from forced circulation to natural circulation flow. Subsequently, during long-term heat removal phase all temperatures evolve very slowly due to the large thermal inertia of the primary and buffer pool systems. The results clearly show that passive safety PRACS can effectively transfer decay heat from the primary system to the buffer pool by natural circulation. The DRACS system in turn can effectively transfer the decay heat to the environment.« less

Critical Design Issues of Tokamak Cooling Water System of ITER's Fusion Reactor

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

Kim, Seokho H; Berry, Jan

U.S. ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). The TCWS transfers heat generated in the Tokamak to cooling water during nominal pulsed operation 850 MW at up to 150 C and 4.2 MPa water pressure. This water contains radionuclides because impurities (e.g., tritium) diffuse from in-vessel components and the vacuum vessel by water baking at 200 240 C at up to 4.4MPa, and corrosion products become activated by neutron bombardment. The system is designated as safety important class (SIC) and will be fabricated to comply with the French Order concerning nuclearmore » pressure equipment (December 2005) and the EU Pressure Equipment Directive using ASME Section VIII, Div 2 design codes. The complexity of the TCWS design and fabrication presents unique challenges. Conceptual design of this one-of-a-kind cooling system has been completed with several issues that need to be resolved to move to next stage of the design. Those issues include flow balancing between over hundreds of branch pipelines in parallel to supply cooling water to blankets, determination of optimum flow velocity while minimizing the potential for cavitation damage, design for freezing protection for cooling water flowing through cryostat (freezing) environment, requirements for high-energy piping design, and electromagnetic impact to piping and components. Although the TCWS consists of standard commercial components such as piping with valves and fittings, heat exchangers, and pumps, complex requirements present interesting design challenges. This paper presents a brief description of TCWS conceptual design and critical design issues that need to be resolved.« less

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

NASA Technical Reports Server (NTRS)

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

1975-01-01

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

Solar photochemical process engineering for production of fuels and chemicals

NASA Technical Reports Server (NTRS)

Biddle, J. R.; Peterson, D. B.; Fujita, T.

1984-01-01

The engineering costs and performance of a nominal 25,000 scmd (883,000 scfd) photochemical plant to produce dihydrogen from water were studied. Two systems were considered, one based on flat-plate collector/reactors and the other on linear parabolic troughs. Engineering subsystems were specified including the collector/reactor, support hardware, field transport piping, gas compression equipment, and balance-of-plant (BOP) items. Overall plant efficiencies of 10.3 and 11.6% are estimated for the flat-plate and trough systems, respectively, based on assumed solar photochemical efficiencies of 12.9 and 14.6%. Because of the opposing effects of concentration ratio and operating temperature on efficiency, it was concluded that reactor cooling would be necessary with the trough system. Both active and passive cooling methods were considered. Capital costs and energy costs, for both concentrating and non-concentrating systems, were determined and their sensitivity to efficiency and economic parameters were analyzed. The overall plant efficiency is the single most important factor in determining the cost of the fuel.

Solar photochemical process engineering for production of fuels and chemicals

NASA Technical Reports Server (NTRS)

Biddle, J. R.; Peterson, D. B.; Fujita, T.

1985-01-01

The engineering costs and performance of a nominal 25,000 scmd (883,000 scfd) photochemical plant to produce dihydrogen from water were studied. Two systems were considered, one based on flat-plate collector/reactors and the other on linear parabolic troughs. Engineering subsystems were specified including the collector/reactor, support hardware, field transport piping, gas compression equipment, and balance-of-plant (BOP) items. Overall plant efficiencies of 10.3 and 11.6 percent are estimated for the flat-plate and trough systems, respectively, based on assumed solar photochemical efficiencies of 12.9 and 14.6 percent. Because of the opposing effects of concentration ratio and operating temperature on efficiency, it was concluded that reactor cooling would be necessary with the trough system. Both active and passive cooling methods were considered. Capital costs and energy costs, for both concentrating and non-concentrating systems, were determined and their sensitivity to efficiency and economic parameters were analyzed. The overall plant efficiency is the single most important factor in determining the cost of the fuel.

Re-weldability tests of irradiated 316L(N) stainless steel using laser welding technique

NASA Astrophysics Data System (ADS)

Yamada, Hirokazu; Kawamura, Hiroshi; Tsuchiya, Kunihiko; Kalinin, George; Kohno, Wataru; Morishima, Yasuo

2002-12-01

SS316L(N)-IG is the candidate material for the in-vessel and ex-vessel components of fusion reactors such as ITER (International Thermonuclear Experimental Reactor). This paper describes a study on re-weldability of un-irradiated and/or irradiated SS316L(N)-IG and the effect of helium generation on the mechanical properties of the weld joint. The laser welding process is used for re-welding of the water cooling branch pipeline repairs. It is clarified that re-welding of SS316L(N)-IG irradiated up to about 0.2 dpa (3.3 appm He) can be carried out without a serious deterioration of tensile properties due to helium accumulation. Therefore, repair of the ITER blanket cooling pipes can be performed by the laser welding process.

Apparatus for controlling molten core debris

DOEpatents

Golden, Martin P. [Trafford, PA; Tilbrook, Roger W. [Monroeville, PA; Heylmun, Neal F. [Pittsburgh, PA

1977-07-19

Apparatus for containing, cooling, diluting, dispersing and maintaining subcritical the molten core debris assumed to melt through the bottom of a nuclear reactor pressure vessel in the unlikely event of a core meltdown. The apparatus is basically a sacrificial bed system which includes an inverted conical funnel, a core debris receptacle including a spherical dome, a spherically layered bed of primarily magnesia bricks, a cooling system of zig-zag piping in graphite blocks about and below the bed and a cylindrical liner surrounding the graphite blocks including a steel shell surrounded by firebrick. Tantalum absorber rods are used in the receptacle and bed.

Non-intrusive cooling system

DOEpatents

Morrison, Edward F.; Bergman, John W.

2001-05-22

A readily replaceable heat exchange cooling jacket for applying fluid to a system conduit pipe. The cooling jacket comprises at least two members, separable into upper and lower portions. A chamber is formed between the conduit pipe and cooling jacket once the members are positioned about the pipe. The upper portion includes a fluid spray means positioned above the pipe and the bottom portion includes a fluid removal means. The heat exchange cooling jacket is adaptable with a drain tank, a heat exchanger, a pump and other standard equipment to provide a system for removing heat from a pipe. A method to remove heat from a pipe, includes the steps of enclosing a portion of the pipe with a jacket to form a chamber between an outside surface of the pipe and the cooling jacket; spraying cooling fluid at low pressure from an upper portion of the cooling jacket, allowing the fluid to flow downwardly by gravity along the surface of the pipe toward a bottom portion of the chamber; and removing the fluid at the bottom portion of the chamber.

Thermal synthesis apparatus

DOEpatents

Fincke, James R [Idaho Falls, ID; Detering, Brent A [Idaho Falls, ID

2009-08-18

An apparatus for thermal conversion of one or more reactants to desired end products includes an insulated reactor chamber having a high temperature heater such as a plasma torch at its inlet end and, optionally, a restrictive convergent-divergent nozzle at its outlet end. In a thermal conversion method, reactants are injected upstream from the reactor chamber and thoroughly mixed with the plasma stream before entering the reactor chamber. The reactor chamber has a reaction zone that is maintained at a substantially uniform temperature. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle, which "freezes" the desired end product(s) in the heated equilibrium reaction stage, or is discharged through an outlet pipe without the convergent-divergent nozzle. The desired end products are then separated from the gaseous stream.

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

Convective cooling in a pool-type research reactor

NASA Astrophysics Data System (ADS)

Sipaun, Susan; Usman, Shoaib

2016-01-01

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

Heat Pipe Solar Receiver for Oxygen Production of Lunar Regolith

NASA Astrophysics Data System (ADS)

Hartenstine, John R.; Anderson, William G.; Walker, Kara L.; Ellis, Michael C.

2009-03-01

A heat pipe solar receiver operating in the 1050° C range is proposed for use in the hydrogen reduction process for the extraction of oxygen from the lunar soil. The heat pipe solar receiver is designed to accept, isothermalize and transfer solar thermal energy to reactors for oxygen production. This increases the available area for heat transfer, and increases throughput and efficiency. The heat pipe uses sodium as the working fluid, and Haynes 230 as the heat pipe envelope material. Initial design requirements have been established for the heat pipe solar receiver design based on information from the NASA In-Situ Resource Utilization (ISRU) program. Multiple heat pipe solar receiver designs were evaluated based on thermal performance, temperature uniformity, and integration with the solar concentrator and the regolith reactor(s). Two designs were selected based on these criteria: an annular heat pipe contained within the regolith reactor and an annular heat pipe with a remote location for the reactor. Additional design concepts have been developed that would use a single concentrator with a single solar receiver to supply and regulate power to multiple reactors. These designs use variable conductance or pressure controlled heat pipes for passive power distribution management between reactors. Following the design study, a demonstration heat pipe solar receiver was fabricated and tested. Test results demonstrated near uniform temperature on the outer surface of the pipe, which will ultimately be in contact with the regolith reactor.

Application of formal optimization techniques in thermal/structural design of a heat-pipe-cooled panel for a hypersonic vehicle

NASA Technical Reports Server (NTRS)

Camarda, Charles J.; Riley, Michael F.

1987-01-01

Nonlinear mathematical programming methods are used to design a radiantly cooled and heat-pipe-cooled panel for a Mach 6.7 transport. The cooled portion of the panel is a hybrid heat-pipe/actively cooled design which uses heat pipes to transport the absorbed heat to the ends of the panel where it is removed by active cooling. The panels are optimized for minimum mass and to satisfy a set of heat-pipe, structural, geometric, and minimum-gage constraints. Two panel concepts are investigated: cylindrical heat pipes embedded in a honeycomb core and an integrated design which uses a web-core heat-pipe sandwich concept. The latter was lighter and resulted in a design which was less than 10 percent heavier than an all actively cooled concept. The heat-pipe concept, however, is redundant and can sustain a single-point failure, whereas the actively cooled concept cannot. An additional study was performed to determine the optimum number of coolant manifolds per panel for a minimum-mass design.

Summary of space nuclear reactor power systems, 1983--1992

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

Buden, D.

1993-08-11

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressedmore » from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.« less

Summary of space nuclear reactor power systems, 1983 - 1992

NASA Astrophysics Data System (ADS)

Buden, D.

1993-08-01

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987-88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

Inherently Safe Fission Power System for Lunar Outposts

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.; El-Genk, Mohamed S.

2013-09-01

This paper presents the Solid Core-Sectored Compact Reactor (SC-SCoRe) and power system for future lunar outposts. The power system nominally provides 38 kWe continuously for 21 years, employs static components and has no single point failures in reactor cooling or power generation. The reactor core has six sectors, each has a separate pair of primary and secondary loops with liquid NaK-56 working fluid, thermoelectric (TE) power conversion and heat-pipes radiator panels. The electromagnetic (EM) pumps in the primary and secondary loops, powered with separate TE power units, ensure operation reliability and passive decay heat removal from the reactor after shutdown. The reactor poses no radiological concerns during launch, and remains sufficiently subcritical, with the radial reflector dissembled, when submerged in wet sand and the core flooded with seawater, following a launch abort accident. After 300 years of storage below grade on the Moon, the total radioactivity in the post-operation reactor drops below 164 Ci, a low enough radioactivity for a recovery and safe handling of the reactor.

Apparatus for controlling molten core debris. [LMFBR

DOEpatents

Golden, M.P.; Tilbrook, R.W.; Heylmun, N.F.

1977-07-19

Disclosed is an apparatus for containing, cooling, diluting, dispersing and maintaining subcritical the molten core debris assumed to melt through the bottom of a nuclear reactor pressure vessel in the unlikely event of a core meltdown. The apparatus is basically a sacrificial bed system which includes an inverted conical funnel, a core debris receptacle including a spherical dome, a spherically layered bed of primarily magnesia bricks, a cooling system of zig-zag piping in graphite blocks about and below the bed and a cylindrical liner surrounding the graphite blocks including a steel shell surrounded by firebrick. Tantalum absorber rods are used in the receptacle and bed. 9 claims, 22 figures.

Shielded fluid stream injector for particle bed reactor

DOEpatents

Notestein, John E.

1993-01-01

A shielded fluid-stream injector assembly is provided for particle bed reactors. The assembly includes a perforated pipe injector disposed across the particle bed region of the reactor and an inverted V-shaped shield placed over the pipe, overlapping it to prevent descending particles from coming into direct contact with the pipe. The pipe and shield are fixedly secured at one end to the reactor wall and slidably secured at the other end to compensate for thermal expansion. An axially extending housing aligned with the pipe and outside the reactor and an in-line reamer are provided for removing deposits from the inside of the pipe. The assembly enables fluid streams to be injected and distributed uniformly into the particle bed with minimized clogging of injector ports. The same design may also be used for extraction of fluid streams from particle bed reactors.

Modelling the radiolysis of RSG-GAS primary cooling water

NASA Astrophysics Data System (ADS)

Butarbutar, S. L.; Kusumastuti, R.; Subekti, M.; Sunaryo, G. R.

2018-02-01

Water chemistry control for light water coolant reactor required a reliable understanding of radiolysis effect in mitigating corrosion and degradation of reactor structure material. It is known that oxidator products can promote the corrosion, cracking and hydrogen pickup both in the core and in the associated piping components of the reactor. The objective of this work is to provide the radiolysis model of RSG GAS cooling water and further more to predict the oxidator concentration which can lead to corrosion of reactor material. Direct observations or measurements of the chemistry in and around the high-flux core region of a nuclear reactor are difficult due to the extreme conditions of high temperature, pressure, and mixed radiation fields. For this reason, chemical models and computer simulations of the radiolysis of water under these conditions are an important route of investigation. FACSIMILE were used to calculate the concentration of O2 formed at relatively long-time by the pure water γ and neutron irradiation (pH=7) at temperature between 25 and 50 °C. This simulation method is based on a complex chemical reaction kinetic. In this present work, 300 MeV-proton were used to mimic γ-rays radiolysis and 2 MeV fast neutrons. Concentration of O2 were calculated at 10-6 - 106 s time scale.

Thermostructural applications of heat pipes for cooling leading edges of high-speed aerospace vehicles

NASA Technical Reports Server (NTRS)

Camarda, Charles J.; Glass, David E.

1992-01-01

Heat pipes have been considered for use on wing leading edge for over 20 years. Early concepts envisioned metal heat pipes cooling a metallic leading edge. Several superalloy/sodium heat pipes were fabricated and successfully tested for wing leading edge cooling. Results of radiant heat and aerothermal testing indicate the feasibility of using heat pipes to cool the stagnation region of shuttle-type space transportation systems. The test model withstood a total seven radiant heating tests, eight aerothermal tests, and twenty-seven supplemental radiant heating tests. Cold-wall heating rates ranged from 21 to 57 Btu/sq ft-s and maximum operating temperatures ranged from 1090 to 1520 F. Follow-on studies investigated the application of heat pipes to cool the stagnation regions of single-stage-to-orbit and advanced shuttle vehicles. Results of those studies indicate that a 'D-shaped' structural design can reduce the mass of the heat-pipe concept by over 44 percent compared to a circular heat-pipe geometry. Simple analytical models for heat-pipe startup from the frozen state (working fluid initially frozen) were adequate to approximate transient, startup, and steady-state heat-pipe performance. Improvement in analysis methods has resulted in the development of a finite-element analysis technique to predict heat-pipe startup from the frozen state. However, current requirements of light-weight design and reliability suggest that metallic heat pipes embedded in a refractory composite material should be used. This concept is the concept presently being evaluated for NASP. A refractory-composite/heat-pipe-cooled wing leading edge is currently being considered for the National Aero-Space Plane (NASP). This concept uses high-temperature refractory-metal/lithium heat pipes embedded within a refractory-composite structure and is significantly lighter than an actively cooled wing leading edge because it eliminates the need for active cooling during ascent and descent. Since the NASP vehicle uses cryogenic hydrogen to cool structural components and then burns this fuel in the combustor, hydrogen necessary for descent cooling only, when the vehicle is unpowered, is considered to be a weight penalty. Details of the design of the refractory-composite/heat-pipe-cooled wing leading edge are currently being investigated. Issues such as thermal contact resistance and thermal stress are also being investigated.

Thermostructural applications of heat pipes for cooling leading edges of high-speed aerospace vehicles

NASA Astrophysics Data System (ADS)

Camarda, Charles J.; Glass, David E.

1992-10-01

Heat pipes have been considered for use on wing leading edge for over 20 years. Early concepts envisioned metal heat pipes cooling a metallic leading edge. Several superalloy/sodium heat pipes were fabricated and successfully tested for wing leading edge cooling. Results of radiant heat and aerothermal testing indicate the feasibility of using heat pipes to cool the stagnation region of shuttle-type space transportation systems. The test model withstood a total seven radiant heating tests, eight aerothermal tests, and twenty-seven supplemental radiant heating tests. Cold-wall heating rates ranged from 21 to 57 Btu/sq ft-s and maximum operating temperatures ranged from 1090 to 1520 F. Follow-on studies investigated the application of heat pipes to cool the stagnation regions of single-stage-to-orbit and advanced shuttle vehicles. Results of those studies indicate that a 'D-shaped' structural design can reduce the mass of the heat-pipe concept by over 44 percent compared to a circular heat-pipe geometry. Simple analytical models for heat-pipe startup from the frozen state (working fluid initially frozen) were adequate to approximate transient, startup, and steady-state heat-pipe performance. Improvement in analysis methods has resulted in the development of a finite-element analysis technique to predict heat-pipe startup from the frozen state. However, current requirements of light-weight design and reliability suggest that metallic heat pipes embedded in a refractory composite material should be used. This concept is the concept presently being evaluated for NASP. A refractory-composite/heat-pipe-cooled wing leading edge is currently being considered for the National Aero-Space Plane (NASP). This concept uses high-temperature refractory-metal/lithium heat pipes embedded within a refractory-composite structure and is significantly lighter than an actively cooled wing leading edge because it eliminates the need for active cooling during ascent and descent. Since the NASP vehicle uses cryogenic hydrogen to cool structural components and then burns this fuel in the combustor, hydrogen necessary for descent cooling only, when the vehicle is unpowered, is considered to be a weight penalty. Details of the design of the refractory-composite/heat-pipe-cooled wing leading edge are currently being investigated. Issues such as thermal contact resistance and thermal stress are also being investigated.

3D-FE Modeling of 316 SS under Strain-Controlled Fatigue Loading and CFD Simulation of PWR Surge Line

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

Mohanty, Subhasish; Barua, Bipul; Listwan, Joseph

In financial year 2017, we are focusing on developing a mechanistic fatigue model of surge line pipes for pressurized water reactors (PWRs). To that end, we plan to perform the following tasks: (1) conduct stress- and strain-controlled fatigue testing of surge-line base metal such as 316 stainless steel (SS) under constant, variable, and random fatigue loading, (2) develop cyclic plasticity material models of 316 SS, (3) develop one-dimensional (1D) analytical or closed-form model to validate the material models and to understand the mechanics associated with 316 SS cyclic hardening and/or softening, (4) develop three-dimensional (3D) finite element (FE) models withmore » implementation of evolutionary cyclic plasticity, and (5) develop computational fluid dynamics (CFD) model for thermal stratification, thermal-mechanical stress, and fatigue of example reactor components, such as a PWR surge line under plant heat-up, cool-down, and normal operation with/without grid-load-following. This semi-annual progress report presents the work completed on the above tasks for a 316 SS laboratory-scale specimen subjected to strain-controlled cyclic loading with constant, variable, and random amplitude. This is the first time that the accurate 3D-FE modeling of the specimen for its entire fatigue life, including the hardening and softening behavior, has been achieved. We anticipate that this work will pave the way for the development of a fully mechanistic-computer model that can be used for fatigue evaluation of safety-critical metallic components, which are traditionally evaluated by heavy reliance on time-consuming and costly test-based approaches. This basic research will not only help the nuclear reactor industry for fatigue evaluation of reactor components in a cost effective and less time-consuming way, but will also help other safety-related industries, such as aerospace, which is heavily dependent on test-based approaches, where a single full-scale fatigue test can cost millions of dollars and require years of effort to conduct. Toward our goal of demonstration of fully mechanistic fatigue evaluation of reactor components, we also started work on developing a component-level computer model of reactor components, such as 316 SS surge line pipe. This requires developing a thermal-mechanical stress analysis model of the reactor surge line, which, in turn, requires time-dependent temperature and stratification information along the boundary of the pipe. Toward that goal, CFD models of surge lines are being developed. In this report, we also present some preliminary results showing the temperature conditions along the surge line wall under reactor heat-up, cool-down, and steady-state power operation.« less

Material Requirements, Selection And Development for the Proposed JIMO SpacePower System

NASA Astrophysics Data System (ADS)

Ring, P. J.; Sayre, E. D.

2004-02-01

NASA is proposing a major new nuclear Space initiative-The Jupiter Icy Moons Orbiter (JIMO). A mission such as this inevitably requires a significant power source both for propulsion and for on-board power. Three reactor concepts, liquid metal cooled, heat pipe cooled and gas cooled are being considered together with three power conversion systems Brayton (cycle), Thermoelectric and Stirling cycles, and possibly Photo voltaics for future systems. Regardless of the reactor system selected it is almost certain that high temperature (materials), refractory alloys, will be required. This paper revisits the material selection options, reviewing the rationale behind the SP-100 selection of Nb-1Zr as the major cladding and structural material and considers the alternatives and developments needed for the longer duty cycle of the JIMO power supply. A side glance is also taken at the basis behind the selection of Uranium nitride fuel over UO2 or UC and a brief discussion of the reason for the selection of Lithium as the liquid metal coolant for SP-100 over other liquid metals.

Convective cooling in a pool-type research reactor

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

Sipaun, Susan, E-mail: susan@nm.gov.my; Usman, Shoaib, E-mail: usmans@mst.edu

2016-01-22

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to passmore » through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.« 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

Report on FY15 alloy 617 code rules development

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

Sham, Sam; Jetter, Robert I; Hollinger, Greg

2015-09-01

Due to its strength at very high temperatures, up to 950°C (1742°F), Alloy 617 is the reference construction material for structural components that operate at or near the outlet temperature of the very high temperature gas-cooled reactors. However, the current rules in the ASME Section III, Division 5 Subsection HB, Subpart B for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 650°C (1200°F) (Corum and Brass, Proceedings of ASME 1991 Pressure Vessels and Piping Conference, PVP-Vol. 215, p.147, ASME, NY, 1991). The rationalemore » for this exclusion is that at higher temperatures it is not feasible to decouple plasticity and creep, which is the basis for the current simplified rules. This temperature, 650°C (1200°F), is well below the temperature range of interest for this material for the high temperature gas-cooled reactors and the very high temperature gas-cooled reactors. The only current alternative is, thus, a full inelastic analysis requiring sophisticated material models that have not yet been formulated and verified. To address these issues, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (EPP) analysis methods applicable to very high temperatures. The proposed rules for strain limits and creep-fatigue evaluation were initially documented in the technical literature (Carter, Jetter and Sham, Proceedings of ASME 2012 Pressure Vessels and Piping Conference, papers PVP 2012 28082 and PVP 2012 28083, ASME, NY, 2012), and have been recently revised to incorporate comments and simplify their application. Background documents have been developed for these two code cases to support the ASME Code committee approval process. These background documents for the EPP strain limits and creep-fatigue code cases are documented in this report.« less

Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

NASA Technical Reports Server (NTRS)

Briggs, Maxwell H.; Gibson, Marc A.; Sanzi, James

2017-01-01

The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one to be used at the Device Assembly Facility (DAF) at the Nevada National Security Site. Fabrication, assembly, and testing of the depleted uranium core has allowed for higher fidelity system level testing at GRC, and has validated the fabrication methods to be used on the highly enriched uranium core that will supply heat for the DAF KRUSTY demonstration.

Cooling Characteristics of an Experimental Tail-pipe Burner with an Annular Cooling-air Passage

NASA Technical Reports Server (NTRS)

Kaufman, Harold R; Koffel, William K

1952-01-01

The effects of tail-pipe fuel-air ratio (exhaust-gas temperatures from approximately 3060 degrees to 3825 degrees R), radial distributiion of tail-pipe fuel flow, and mass flow of combustion gas and the inside wall were determined for an experimental tail-pipe burner cooled by air flowing through and insulated cooling-air to combustion gas mass flow from 0.066 to 0.192 were also determined.

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

Heat Transfer Modeling of an Annular On-Line Spray Water Cooling Process for Electric-Resistance-Welded Steel Pipe

PubMed Central

Chen, Zejun; Han, Huiquan; Ren, Wei; Huang, Guangjie

2015-01-01

On-line spray water cooling (OSWC) of electric-resistance-welded (ERW) steel pipes can replace the conventional off-line heat treatment process and become an important and critical procedure. The OSWC process improves production efficiency, decreases costs, and enhances the mechanical properties of ERW steel pipe, especially the impact properties of the weld joint. In this paper, an annular OSWC process is investigated based on an experimental simulation platform that can obtain precise real-time measurements of the temperature of the pipe, the water pressure and flux, etc. The effects of the modes of annular spray water cooling and related cooling parameters on the mechanical properties of the pipe are investigated. The temperature evolutions of the inner and outer walls of the pipe are measured during the spray water cooling process, and the uniformity of mechanical properties along the circumferential and longitudinal directions is investigated. A heat transfer coefficient model of spray water cooling is developed based on measured temperature data in conjunction with simulation using the finite element method. Industrial tests prove the validity of the heat transfer model of a steel pipe undergoing spray water cooling. The research results can provide a basis for the industrial application of the OSWC process in the production of ERW steel pipes. PMID:26201073

Heat Transfer Modeling of an Annular On-Line Spray Water Cooling Process for Electric-Resistance-Welded Steel Pipe.

PubMed

Chen, Zejun; Han, Huiquan; Ren, Wei; Huang, Guangjie

2015-01-01

On-line spray water cooling (OSWC) of electric-resistance-welded (ERW) steel pipes can replace the conventional off-line heat treatment process and become an important and critical procedure. The OSWC process improves production efficiency, decreases costs, and enhances the mechanical properties of ERW steel pipe, especially the impact properties of the weld joint. In this paper, an annular OSWC process is investigated based on an experimental simulation platform that can obtain precise real-time measurements of the temperature of the pipe, the water pressure and flux, etc. The effects of the modes of annular spray water cooling and related cooling parameters on the mechanical properties of the pipe are investigated. The temperature evolutions of the inner and outer walls of the pipe are measured during the spray water cooling process, and the uniformity of mechanical properties along the circumferential and longitudinal directions is investigated. A heat transfer coefficient model of spray water cooling is developed based on measured temperature data in conjunction with simulation using the finite element method. Industrial tests prove the validity of the heat transfer model of a steel pipe undergoing spray water cooling. The research results can provide a basis for the industrial application of the OSWC process in the production of ERW steel pipes.

Heat-Pipe-Cooled Leading Edges for Hypersonic Vehicles

NASA Technical Reports Server (NTRS)

Glass, David E.

2006-01-01

Heat pipes can be used to effectively cool wing leading edges of hypersonic vehicles. . Heat-pipe leading edge development. Design validation heat pipe testing confirmed design. Three heat pipes embedded and tested in C/C. Single J-tube heat pipe fabricated and testing initiated. HPCLE work is currently underway at several locations.

DynMo: Dynamic Simulation Model for Space Reactor Power Systems

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed; Tournier, Jean-Michel

2005-02-01

A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK® platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed.

Heat pipe cooled power magnetics

NASA Technical Reports Server (NTRS)

Chester, M. S.

1979-01-01

A high frequency, high power, low specific weight (0.57 kg/kW) transformer developed for space use was redesigned with heat pipe cooling allowing both a reduction in weight and a lower internal temperature rise. The specific weight of the heat pipe cooled transformer was reduced to 0.4 kg/kW and the highest winding temperature rise was reduced from 40 C to 20 C in spite of 10 watts additional loss. The design loss/weight tradeoff was 18 W/kg. Additionally, allowing the same 40 C winding temperature rise as in the original design, the KVA rating is increased to 4.2 KVA, demonstrating a specific weight of 0.28 kg/kW with the internal loss increased by 50W. This space environment tested heat pipe cooled design performed as well electrically as the original conventional design, thus demonstrating the advantages of heat pipes integrated into a high power, high voltage magnetic. Another heat pipe cooled magnetic, a 3.7 kW, 20A input filter inductor was designed, developed, built, tested, and described. The heat pipe cooled magnetics are designed to be Earth operated in any orientation.

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

Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.

The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and implementation of a virtual reactivity feedback control loop; testing and thermal hydraulic characterization of the coolant flow paths for two space reactor concepts; and analysis of heat pipe operation during start-up and steady state operation.

Hydrogen Assisted Cracking and Corrosion of Some Highly Corrosion Resistant Alloys

DTIC Science & Technology

1990-01-01

Stainless Steel", June 1985, and "On the Roles of Corrosion Products in Local Cell Processes", January 1986. Research on the latter has occurred in the...concern. In closed systems. howevter, such as nuclear reactor cooling pipes. acid container systems, fuel cells, and so on. the production of ti, gas and...mernhra lie is also imiportant. fihe stirf.ice should he flat. m-e1I-polished and free of filims. (Whde or other corrosion product film-. :Are easil% formed

Heat pipe cooling for scramjet engines

NASA Technical Reports Server (NTRS)

Silverstein, Calvin C.

1986-01-01

Liquid metal heat pipe cooling systems have been investigated for the combustor liner and engine inlet leading edges of scramjet engines for a missile application. The combustor liner is cooled by a lithium-TZM molybdenum annular heat pipe, which incorporates a separate lithium reservoir. Heat is initially absorbed by the sensible thermal capacity of the heat pipe and liner, and subsequently by the vaporization and discharge of lithium to the atmosphere. The combustor liner temperature is maintained at 3400 F or less during steady-state cruise. The engine inlet leading edge is fabricated as a sodium-superalloy heat pipe. Cooling is accomplished by radiation of heat from the aft surface of the leading edge to the atmosphere. The leading edge temperature is limited to 1700 F or less. It is concluded that heat pipe cooling is a viable method for limiting scramjet combustor liner and engine inlet temperatures to levels at which structural integrity is greatly enhanced.

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

Visualization of Flow in Pressurizer Spray Line Piping and Estimation of Thermal Stress Fluctuation Caused by Swaying of Water Surface

NASA Astrophysics Data System (ADS)

Oumaya, Toru; Nakamura, Akira; Onojima, Daisuke; Takenaka, Nobuyuki

The pressurizer spray line of PWR plants cools reactor coolant by injecting water into pressurizer. Since the continuous spray flow rate during commercial operation of the plant is considered insufficient to fill the pipe completely, there is a concern that a water surface exists in the pipe and may periodically sway. In order to identify the flow regimes in spray line piping and assess their impact on pipe structure, a flow visualization experiment was conducted. In the experiment, air was used substituted for steam to simulate the gas phase of the pressurizer, and the flow instability causing swaying without condensation was investigated. With a full-scale mock-up made of acrylic, flow under room temperature and atmospheric pressure conditions was visualized, and possible flow regimes were identified based on the results of the experiment. Three representative patterns of swaying of water surface were assumed, and the range of thermal stress fluctuation, when the surface swayed instantaneously, was calculated. With the three patterns of swaying assumed based on the visualization experiment, it was confirmed that the thermal stress amplitude would not exceed the fatigue endurance limit prescribed in the Japanese Design and Construction Code.

Modeling of High Capacity Passive Cooling System

DTIC Science & Technology

2009-03-01

Pulsating Heat Pipes : Closed Loop Pulsating Heat Pipes , which is also known as Meandering Capillary Tube Heat Pipe or Closed Loop Oscillating Heat ... Pipe , has emerged in the recent years as a new electronics cooling technology. The Pulsating Heat Pipe is an innovating technology that has gained...horizontal orientation, the operating temperatures are lower. Pulsating heat pipes are capable of higher heat

PROCESS WATER BUILDING, TRA605. AERIAL TAKEN WHILE SEVERAL PIPE TRENCHES ...

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

PROCESS WATER BUILDING, TRA-605. AERIAL TAKEN WHILE SEVERAL PIPE TRENCHES REMAINED OPEN. CAMERA FACES EASTERLY. NOTE DUAL PIPES BETWEEN REACTOR BUILDING AND NORTH SIDE OF PROCESS WATER BUILDING. PIPING NEAR WORKING RESERVOIR HEADS FOR RETENTION RESERVOIR. PIPE FROM DEMINERALIZER ENTERS MTR FROM NORTH. SEE ALSO TRENCH FOR COOLANT AIR DUCT AT SOUTH SIDE OF MTR AND LEADING TO FAN HOUSE AND STACK. INL NEGATIVE NO. 2966-A. Unknown Photographer, 7/31/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Closed Form Equations for the Preliminary Design of a Heat-Pipe-Cooled Leading Edge

NASA Technical Reports Server (NTRS)

Glass, David E.

1998-01-01

A set of closed form equations for the preliminary evaluation and design of a heat-pipe-cooled leading edge is presented. The set of equations can provide a leading-edge designer with a quick evaluation of the feasibility of using heat-pipe cooling. The heat pipes can be embedded in a metallic or composite structure. The maximum heat flux, total integrated heat load, and thermal properties of the structure and heat-pipe container are required input. The heat-pipe operating temperature, maximum surface temperature, heat-pipe length, and heat pipe-spacing can be estimated. Results using the design equations compared well with those from a 3-D finite element analysis for both a large and small radius leading edge.

Water inventory management in condenser pool of boiling water reactor

DOEpatents

Gluntz, Douglas M.

1996-01-01

An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

Water inventory management in condenser pool of boiling water reactor

DOEpatents

Gluntz, D.M.

1996-03-12

An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

Cool-down and frozen start-up behavior of a grooved water heat pipe

NASA Technical Reports Server (NTRS)

Jang, Jong Hoon

1990-01-01

A grooved water heat pipe was tested to study its characteristics during the cool-down and start-up periods. The water heat pipe was cooled down from the ambient temperature to below the freezing temperature of water. During the cool-down, isothermal conditions were maintained at the evaporator and adiabatic sections until the working fluid was frozen. When water was frozen along the entire heat pipe, the heat pipe was rendered inactive. The start-up of the heat pipe from this state was studied under several different operating conditions. The results show the existence of large temperature gradients between the evaporator and the condenser, and the moving of the melting front of the working fluid along the heat pipe. Successful start-up was achieved for some test cases using partial gravity assist. The start-up behavior depended largely on the operating conditions.

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.

Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINK{sup TM}

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

Wright, Steven A.; Sanchez, Travis

2005-02-06

The operation of space reactors for both in-space and planetary operations will require unprecedented levels of autonomy and control. Development of these autonomous control systems will require dynamic system models, effective control methodologies, and autonomous control logic. This paper briefly describes the results of reactor, power-conversion, and control models that are implemented in SIMULINK{sup TM} (Simulink, 2004). SIMULINK{sup TM} is a development environment packaged with MatLab{sup TM} (MatLab, 2004) that allows the creation of dynamic state flow models. Simulation modules for liquid metal, gas cooled reactors, and electrically heated systems have been developed, as have modules for dynamic power-conversion componentsmore » such as, ducting, heat exchangers, turbines, compressors, permanent magnet alternators, and load resistors. Various control modules for the reactor and the power-conversion shaft speed have also been developed and simulated. The modules are compiled into libraries and can be easily connected in different ways to explore the operational space of a number of potential reactor, power-conversion system configurations, and control approaches. The modularity and variability of these SIMULINK{sup TM} models provides a way to simulate a variety of complete power generation systems. To date, both Liquid Metal Reactors (LMR), Gas Cooled Reactors (GCR), and electric heaters that are coupled to gas-dynamics systems and thermoelectric systems have been simulated and are used to understand the behavior of these systems. Current efforts are focused on improving the fidelity of the existing SIMULINK{sup TM} modules, extending them to include isotopic heaters, heat pipes, Stirling engines, and on developing state flow logic to provide intelligent autonomy. The simulation code is called RPC-SIM (Reactor Power and Control-Simulator)« less

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

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

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

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.

Split-core heat-pipe reactors for out-of-pile thermionic power systems.

NASA Technical Reports Server (NTRS)

Niederauer, G.; Lantz, E.; Breitweiser, R.

1971-01-01

Description of the concept of splitting a heat-pipe reactor for out-of-core thermionics into two identical halves and using the resulting center gap for reactivity control. Short Li-W reactor heat pipes penetrate the axial reflectors and form a heat exchanger with long heat pipes which wind through the shield to the thermionic diodes. With one reactor half anchored to the shield, the other is attached to a long arm with a pivot behind the shield and swings through a small arc for reactivity control. A safety shim prevents large reactivity inputs, and a fueled control arm drive shaft acts as a power stabilizer. Reactors fueled with U-235C and with U-233C have been studied.-

Application of heat pipe technology in permanent mold casting of nonferrous alloys

NASA Astrophysics Data System (ADS)

Elalem, Kaled

The issue of mold cooling is one, which presents a foundry with a dilemma. On the one hand; the use of air for cooling is safe and practical, however, it is not very effective and high cost. On the other hand, water-cooling can be very effective but it raises serious concerns about safety, especially with a metal such as magnesium. An alternative option that is being developed at McGill University uses heat pipe technology to carry out the cooling. The experimental program consisted of designing a permanent mold to produce AZ91E magnesium alloy and A356 aluminum alloy castings with shrinkage defects. Heat pipes were then used to reduce these defects. The heat pipes used in this work are novel and are patent pending. They are referred to as McGill Heat Pipes. Computer modeling was used extensively in designing the mold and the heat pipes. Final designs for the mold and the heat pipes were chosen based on the modeling results. Laboratory tests of the heat pipe were performed before conducting the actual experimental plan. The laboratory testing results verified the excellent performance of the heat pipes as anticipated by the model. An industrial mold made of H13 tool steel was constructed to cast nonferrous alloys. The heat pipes were installed and initial testing and actual industrial trials were conducted. This is the first time where a McGill heat pipe was used in an industrial permanent mold casting process for nonferrous alloys. The effects of cooling using heat pipes on AZ91E and A356 were evaluated using computer modeling and experimental trials. Microstructural analyses were conducted to measure the secondary dendrite arm spacing, SDAS, and the grain size to evaluate the cooling effects on the castings. The modeling and the experimental results agreed quite well. The metallurgical differences between AZ91E and A356 were investigated using modeling and experimental results. Selected results from modeling, laboratory and industrial trials are presented. The results show a promising future for heat pipe technology in cooling permanent molds for the casting of nonferrous alloys.

Non-Intrusive Velocity Measurements with MTV During DCC Event in the HTTF

NASA Technical Reports Server (NTRS)

Andre, M. A.; Bardet, P. M.; Cadell, S. R.; Woods, B.; Burns, R. A.; Danehy, P. M.

2017-01-01

Velocity profiles are measured using molecular tagging velocimetry (MTV) in the high temperature test facility (HTTF) at Oregon State University during a depressurized conduction cooldown (DCC) event. The HTTF is a quarter scale electrically heated nuclear reactor simulator designed to replicate various accident scenarios. During a DCC, a double ended guillotine break results in the reactor pressure vessel (RPV) depressurizing into the reactor cavity and ultimately leading to air ingress in the reactor core (lock-exchange and gas diffusion). It is critical to understand the resulting buoyancy-driven flow to characterize the reactor self-cooling capacity through natural circulation. During tests conducted at ambient pressure and temperature, the RPV containing helium is opened (via the hot and cold legs) to a large vessel filled with nitrogen to simulate the atmosphere. The velocity profile on the hot leg pipe centerline is recorded at 10 Hz with MTV based on NO tracers. The precision of the velocimetry was measured to be 0.02 m/s in quiescent flow prior to the tests. A helium flow from the RPV is initially observed in the top quarter of the pipe. During the first 20 seconds of the event, helium flows out of the RPV with a maximum velocity below 2 m/s. The velocity profile transitions from parabolic to linear in character and decays slowly over the rest of the recording; peak velocities of 0.2 m/s are observed after 30 min. A counter-flow of nitrogen is also observed intermittently, which occurs at lower velocities (>0.1 m/s).

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.; Morton, T. J.

2006-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG testing will utilize a higher fidelity point kinetics model to control core power transients, and reactivity feedback will be based on localized feedback coefficients and several independent temperature measurements taken within the core block. This paper presents preliminary test results and discusses the methodology that will be implemented in follow-on DDG testing and the additional instrumentation required to implement high fidelity dynamic testing.

Best-estimate coupled RELAP/CONTAIN analysis of inadvertent BWR ADS valve opening transient

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

Feltus, M.A.; Muftuoglu, A.K.

1993-01-01

Noncondensible gases may become dissolved in boiling water reactor (BWR) water-level instrumentation during normal operations. Any dissolved noncondensible gases inside these water columns may come out of solution during rapid depressurization events and displace water from the reference leg piping, resulting in a false high level. Significant errors in water-level indication are not expected to occur until the reactor pressure vessel (RPV) pressure has dropped below [approximately]450 psig. These water level errors may cause a delay or failure in emergency core cooling system (ECCS) actuation. The RPV water level is monitored using the pressure of a water column having amore » varying height (reactor water level) that is compared to the pressure of a water column maintained at a constant height (reference level). The reference legs have small-diameter pipes with varying lengths that provide a constant head of water and are located outside the drywell. The amount of noncondensible gases dissolved in each reference leg is very dependent on the amount of leakage from the reference leg and its geometry and interaction of the reactor coolant system with the containment, i.e., torus or suppression pool, and reactor building. If a rapid depressurization causes an erroneously high water level, preventing automatic ECCS actuation, it becomes important to determine if there would be other adequate indications for operator response. In the postulated inadvertent opening of all seven automatic depressurization system (ADS) valves, the ECCS signal on high drywell pressure would be circumvented because the ADS valves discharge directly into the suppression pool. A best-estimate analysis of such an inadvertent opening of all ADS valves would have to consider the thermal-hydraulic coupling between the pool, drywell, reactor building, and RPV.« less

Fabrication and Testing of Mo-Re Heat Pipes Embedded in Carbon/Carbon

NASA Technical Reports Server (NTRS)

Glass, David E.; Merrigan, Michael A.; Sena, J. Tom

1998-01-01

Refractory-composite/heat-pipe-cooled wing an tail leading edges are being considered for use on hypersonic vehicles to limit maximum temperatures to values below material reuse limits and to eliminate the need to actively cool the leading edges. The development of a refractory-composite/heat-pipe-cooled leading edge has evolved from the design stage to the fabrication and testing of heat pipes embedded in carbon/carbon (C/C). A three-foot-long, molybdenum-rhenium heat pipe with a lithium working fluid was fabricated and tested at an operating temperature of 2460 F to verify the individual heat-pipe design. Following the fabrication of this heat pipe, three additional heat pipes were fabricated and embedded in C/C. The C/C heat-pipe test article was successfully tested using quartz lamps in a vacuum chamber in both a horizontal and vertical orientation. Start up and steady state data are presented for the C/C heat-pipe test article. Radiography and eddy current evaluations were performed on the test article.

Experimental Investigation of Thermal Performance of Miniature Heat Pipe Using SiO2-Water Nanofluids.

PubMed

Niu, Yan-Fang; Zhao, Wei-Lin; Gong, Yu-Ying

2015-04-01

The four miniature heat pipes filled with DI water and SiO2-water nanofluids containing different volume concentrations (0.2%, 0.6% and 1.0%) are experimentally measured on the condition of air and water cooling. The wall temperature and the thermal resistance are investigated for three inclination angles. At the same of inlet heat water temperature in the heat system, it is observed that the total wall temperatures on the evaporator section are almost retaining constant by air cooling and the wall temperatures at the front end of the evaporator section are slightly reduced by water cooling. However, the wall temperatures at the condenser section using SiO2-water nanofluids are all higher than that for DI water on the two cooling conditions. As compared with the heat pipe using DI water, the decreasing of the thermal resistance in heat pipe using nanofluids is about 43.10%-74.46% by air cooling and 51.43%-72.22% by water cooling. These indicate that the utilization of SiO2-water nanofluids as working fluids enhances the performance of the miniature heat pipe. When the four miniature heat pipes are cut to examine at the end of the experiment, a thin coating on the surface of the screen mesh of the heat pipe using SiO2-water nanofluids is found. This may be one reason for reinforcing the heat transfer performance of the miniature heat pipe.

46 CFR 119.425 - Engine exhaust cooling.

Code of Federal Regulations, 2010 CFR

2010-10-01

..., all engine exhaust pipes must be water cooled. (1) Vertical dry exhaust pipes are permissible if installed in compliance with §§ 116.405(c) and 116.970 of this chapter. (2) Horizontal dry exhaust pipes are...) They are installed in compliance with §§ 116.405(c) and 116.970 of this chapter. (b) The exhaust pipe...

Centrifuge Testing of a Partially-Confined FC-72 Spray

DTIC Science & Technology

2006-11-01

induced body forces. Heat transfer associated with closed - loop spray cooling will be affected by acceleration body forces, the extent of which is not...impingement cooling, spray cooling, heat pipes , loop heat pipes , carbon foam impregnated with phase-change materials, and combinations of the above...reduced gravity and elevated gravity experiments to help prove viability of pulsating heat pipes (PHPs) for space applications. The PHPs, filled

Heat Pipes Cool Power Magnetics

NASA Technical Reports Server (NTRS)

Hansen, I.; Chester, M.; Luedke, E.

1983-01-01

Configurations originally developed for space use are effective in any orientation. Heat pipes integrated into high-power, high-frequency, highvoltage spaceflight magnetics reduce weight and improve reliability by lowering internal tempertures. Two heat pipes integrated in design of power transformer cool unit in any orientation. Electrostatic shield conducts heat from windings to heat pipe evaporator. Technology allows dramatic reductions in size and weight, while significantly improving reliability. In addition, all attitude design of heat pipes allows operation of heat pipes independent of local gravity forces.

A feasibility study of heat-pipe-cooled leading edges for hypersonic cruise aircraft

NASA Technical Reports Server (NTRS)

Silverstein, C. C.

1971-01-01

A theoretical study of the use of heat pipe structures for cooling the leading edges of hypersonic cruise aircraft was carried out over a Mach number range of 6 to 12. Preliminary design studies showed that a heat pipe cooling structure with a 33-in. chordwise length could maintain the maximum temperature of a 65 deg sweepback wing with a 0.5-in. leading edge radius below 1600 F during cruise at Mach 8. A few relatively minor changes in the steady-state design of the structure were found necessary to insure satisfactory cooling during the climb to cruise speed and altitude. It was concluded that heat pipe cooling is an attractive, feasible technique for limiting leading edge temperatures of hypersonic cruise aircraft.

46 CFR 119.425 - Engine exhaust cooling.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 46 Shipping 4 2011-10-01 2011-10-01 false Engine exhaust cooling. 119.425 Section 119.425 Shipping... Machinery Requirements § 119.425 Engine exhaust cooling. (a) Except as otherwise provided in this paragraph, all engine exhaust pipes must be water cooled. (1) Vertical dry exhaust pipes are permissible if...

46 CFR 119.425 - Engine exhaust cooling.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 46 Shipping 4 2012-10-01 2012-10-01 false Engine exhaust cooling. 119.425 Section 119.425 Shipping... Machinery Requirements § 119.425 Engine exhaust cooling. (a) Except as otherwise provided in this paragraph, all engine exhaust pipes must be water cooled. (1) Vertical dry exhaust pipes are permissible if...

46 CFR 119.425 - Engine exhaust cooling.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 46 Shipping 4 2014-10-01 2014-10-01 false Engine exhaust cooling. 119.425 Section 119.425 Shipping... Machinery Requirements § 119.425 Engine exhaust cooling. (a) Except as otherwise provided in this paragraph, all engine exhaust pipes must be water cooled. (1) Vertical dry exhaust pipes are permissible if...

46 CFR 119.425 - Engine exhaust cooling.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 46 Shipping 4 2013-10-01 2013-10-01 false Engine exhaust cooling. 119.425 Section 119.425 Shipping... Machinery Requirements § 119.425 Engine exhaust cooling. (a) Except as otherwise provided in this paragraph, all engine exhaust pipes must be water cooled. (1) Vertical dry exhaust pipes are permissible if...

An experimental study of heat pipe thermal management system with wet cooling method for lithium ion batteries

NASA Astrophysics Data System (ADS)

Zhao, Rui; Gu, Junjie; Liu, Jie

2015-01-01

An effective battery thermal management (BTM) system is required for lithium-ion batteries to ensure a desirable operating temperature range with minimal temperature gradient, and thus to guarantee their high efficiency, long lifetime and great safety. In this paper, a heat pipe and wet cooling combined BTM system is developed to handle the thermal surge of lithium-ion batteries during high rate operations. The proposed BTM system relies on ultra-thin heat pipes which can efficiently transfer the heat from the battery sides to the cooling ends where the water evaporation process can rapidly dissipate the heat. Two sized battery packs, 3 Ah and 8 Ah, with different lengths of cooling ends are used and tested through a series high-intensity discharges in this study to examine the cooling effects of the combined BTM system, and its performance is compared with other four types of heat pipe involved BTM systems and natural convection cooling method. A combination of natural convection, fan cooling and wet cooling methods is also introduced to the heat pipe BTM system, which is able to control the temperature of battery pack in an appropriate temperature range with the minimum cost of energy and water spray.

I-NERI Quarterly Technical Report (April 1 to June 30, 2005)

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

Chang Oh; Prof. Hee Cheon NO; Prof. John Lee

2005-06-01

The objective of this Korean/United States/laboratory/university collaboration is to develop new advanced computational methods for safety analysis codes for very-high-temperature gas-cooled reactors (VHTGRs) and numerical and experimental validation of these computer codes. This study consists of five tasks for FY-03: (1) development of computational methods for the VHTGR, (2) theoretical modification of aforementioned computer codes for molecular diffusion (RELAP5/ATHENA) and modeling CO and CO2 equilibrium (MELCOR), (3) development of a state-of-the-art methodology for VHTGR neutronic analysis and calculation of accurate power distributions and decay heat deposition rates, (4) reactor cavity cooling system experiment, and (5) graphite oxidation experiment. Second quartermore » of Year 3: (A) Prof. NO and Kim continued Task 1. As a further plant application of GAMMA code, we conducted two analyses: IAEA GT-MHR benchmark calculation for LPCC and air ingress analysis for PMR 600MWt. The GAMMA code shows comparable peak fuel temperature trend to those of other country codes. The analysis results for air ingress show much different trend from that of previous PBR analysis: later onset of natural circulation and less significant rise in graphite temperature. (B) Prof. Park continued Task 2. We have designed new separate effect test device having same heat transfer area and different diameter and total number of U-bands of air cooling pipe. New design has smaller pressure drop in the air cooling pipe than the previous one as designed with larger diameter and less number of U-bands. With the device, additional experiments have been performed to obtain temperature distributions of the water tank, the surface and the center of cooling pipe on axis. The results will be used to optimize the design of SNU-RCCS. (C) Prof. NO continued Task 3. The experimental work of air ingress is going on without any concern: With nuclear graphite IG-110, various kinetic parameters and reaction rates for the C/CO2 reaction were measured. Then, the rates of C/CO2 reaction were compared to the ones of C/O2 reaction. The rate equation for C/CO2 has been developed. (D) INL added models to RELAP5/ATHENA to cacilate the chemical reactions in a VHTR during an air ingress accident. Limited testing of the models indicate that they are calculating a correct special distribution in gas compositions. (E) INL benchmarked NACOK natural circulation data. (F) Professor Lee et al at the University of Michigan (UM) Task 5. The funding was received from the DOE Richland Office at the end of May and the subcontract paperwork was delivered to the UM on the sixth of June. The objective of this task is to develop a state of the art neutronics model for determining power distributions and decay heat deposition rates in a VHTGR core. Our effort during the reporting period covered reactor physics analysis of coated particles and coupled nuclear-thermal-hydraulic (TH) calculations, together with initial calculations for decay heat deposition rates in the core.« less

Deposition of fission and activation products after the Fukushima Dai-ichi nuclear power plant accident.

PubMed

Shozugawa, Katsumi; Nogawa, Norio; Matsuo, Motoyuki

2012-04-01

The Great Eastern Japan Earthquake on March 11, 2011, damaged reactor cooling systems at Fukushima Dai-ichi nuclear power plant. The subsequent venting operation and hydrogen explosion resulted in a large radioactive nuclide emission from reactor containers into the environment. Here, we collected environmental samples such as soil, plant species, and water on April 10, 2011, in front of the power plant main gate as well as 35 km away in Iitate village, and observed gamma-rays with a Ge(Li) semiconductor detector. We observed activation products ((239)Np and (59)Fe) and fission products ((131)I, (134)Cs ((133)Cs), (137)Cs, (110m)Ag ((109)Ag), (132)Te, (132)I, (140)Ba, (140)La, (91)Sr, (91)Y, (95)Zr, and (95)Nb). (239)Np is the parent nuclide of (239)Pu; (59)Fe are presumably activation products of (58)Fe obtained by corrosion of cooling pipes. The results show that these activation and fission products, diffused within a month of the accident. Copyright © 2012 Elsevier Ltd. All rights reserved.

Simulation of Radioactive Corrosion Product in Primary Cooling System of Japanese Sodium-Cooled Fast Breeder Reactor

NASA Astrophysics Data System (ADS)

Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

Radioactive Corrosion Product (CP) is a main cause of personal radiation exposure during maintenance with no breached fuel in fast breeder reactor (FBR) plants. The most important CP is 54Mn and 60Co. In order to establish techniques of radiation dose estimation for radiation workers in radiation-controlled areas of the FBR, the PSYCHE (Program SYstem for Corrosion Hazard Evaluation) code was developed. We add the Particle Model to the conventional PSYCHE analytical model. In this paper, we performed calculation of CP transfer in JOYO using an improved calculation code in which the Particle Model was added to the PSYCHE. The C/E (calculated / experimentally observed) value for CP deposition was improved through use of this improved PSYCHE incorporating the Particle Model. Moreover, among the percentage of total radioactive deposition accounted for by CP in particle form, 54Mn was estimated to constitute approximately 20 % and 60Co approximately 40 % in the cold-leg region. These calculation results are consistent with the measured results for the actual cold-leg piping in the JOYO.

Fundamental Research on Heat Transfer Characteristics in Shell & Tube Type Ice Forming Cold Energy Storage

NASA Astrophysics Data System (ADS)

Saito, Akio; Utaka, Yoshio; Okawa, Seiji; Ishibashi, Hiroaki

Investigation of heat transfer characteristics in an ice making cold energy storage using a set of horizontal cooling pipes was carried out experimentally. Cooling pipe arrangement, number of pipes used and initial water temperature were varied, and temperature distribution in the tank and the volume of ice formed around the pipe were measured. Natural convection was also observed visually. During the experiment, two kinds of layers were observed. One is the layer where ice forming is interfered by natural convection and its temperature decreases rapidly with an almost uniform temperature distribution, and the other is the layer where ice forms steadily under a stagnant water condition. The former was called that the layer is under a cooling process and the latter that the layer is under an ice forming process. The effect of the experimental parameters, such as the arrangement of the cooling pipes, the number of pipes, the initial water temperature and the flow rate of the cooling medium, on the cooling process and the ice forming process were discussed. Approximate analysis was also carried out and compared with the experimental results. Finally, the relationship between the ice packing factor, which is significant in preventing the blockade, and experimental parameters was discussed.

Hypersonic aerospace vehicle leading edge cooling using heat pipe, transpiration and film cooling techniques

NASA Astrophysics Data System (ADS)

Modlin, James Michael

An investigation was conducted to study the feasibility of cooling hypersonic vehicle leading edge structures exposed to severe aerodynamic surface heat fluxes using a combination of liquid metal heat pipes and surface mass transfer cooling techniques. A generalized, transient, finite difference based hypersonic leading edge cooling model was developed that incorporated these effects and was demonstrated on an assumed aerospace plane-type wing leading edge section and a SCRAMJET engine inlet leading edge section. The hypersonic leading edge cooling model was developed using an existing, experimentally verified heat pipe model. Two applications of the hypersonic leading edge cooling model were examined. An assumed aerospace plane-type wing leading edge section exposed to a severe laminar, hypersonic aerodynamic surface heat flux was studied. A second application of the hypersonic leading edge cooling model was conducted on an assumed one-quarter inch nose diameter SCRAMJET engine inlet leading edge section exposed to both a transient laminar, hypersonic aerodynamic surface heat flux and a type 4 shock interference surface heat flux. The investigation led to the conclusion that cooling leading edge structures exposed to severe hypersonic flight environments using a combination of liquid metal heat pipe, surface transpiration, and film cooling methods appeared feasible.

Results from the decontamination of and the shielding arrangements in the reactor pressure vessel in Oskarshamn 1-1994

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

Lowendahl, B.

1995-03-01

In September 1992 Oskarshamn 1 was shut down in order to carry out measures to correct discovered deficiencies in the emergency cooling systems. Due to the results of a comprehensive non destructive test programme it was decided to perform a major replacement of pipes in the primary systems including a full system decontamination using the Siemens CORD process. The paper briefly presents the satisfying result of the decontamination performed in May-June 1993. When in late June 1993 cracks also were detected in the feed-water pipes situated inside the reactor pressure vessel (RPV) the plans were reconsidered and a large projectmore » was formed with the aim, in a first phase, to verify the integrity of the RPV. In order to make it possible to perform work manually inside the RPV special radiation protection measures had to be carried out. In January 1994 the lower region of the RPV was decontaminated, again using the CORD-process, followed by the installation of a special shielding construction in the RPV. The surprisingly good results of these efforts are also briefly described in the paper.« less

The study and development of the empirical correlations equation of natural convection heat transfer on vertical rectangular sub-channels

NASA Astrophysics Data System (ADS)

Kamajaya, Ketut; Umar, Efrizon; Sudjatmi, K. S.

2012-06-01

This study focused on natural convection heat transfer using a vertical rectangular sub-channel and water as the coolant fluid. To conduct this study has been made pipe heaters are equipped with thermocouples. Each heater is equipped with five thermocouples along the heating pipes. The diameter of each heater is 2.54 cm and 45 cm in length. The distance between the central heating and the pitch is 29.5 cm. Test equipment is equipped with a primary cooling system, a secondary cooling system and a heat exchanger. The purpose of this study is to obtain new empirical correlations equations of the vertical rectangular sub-channel, especially for the natural convection heat transfer within a bundle of vertical cylinders rectangular arrangement sub-channels. The empirical correlation equation can support the thermo-hydraulic analysis of research nuclear reactors that utilize cylindrical fuel rods, and also can be used in designing of baffle-free vertical shell and tube heat exchangers. The results of this study that the empirical correlation equations of natural convection heat transfer coefficients with rectangular arrangement is Nu = 6.3357 (Ra.Dh/x)0.0740.

Off-axis cooling of rotating devices using a crank-shaped heat pipe

DOEpatents

Jankowski, Todd A.; Prenger, F. Coyne; Waynert, Joseph A.

2007-01-30

The present invention is a crank-shaped heat pipe for cooling rotating machinery and a corresponding method of manufacture. The crank-shaped heat pipe comprises a sealed cylindrical tube with an enclosed inner wick structure. The crank-shaped heat pipe includes a condenser section, an adiabatic section, and an evaporator section. The crank-shape is defined by a first curve and a second curve existing in the evaporator section or the adiabatic section of the heat pipe. A working fluid within the heat pipe provides the heat transfer mechanism.

Characterization of a high performance ultra-thin heat pipe cooling module for mobile hand held electronic devices

NASA Astrophysics Data System (ADS)

Ahamed, Mohammad Shahed; Saito, Yuji; Mashiko, Koichi; Mochizuki, Masataka

2017-11-01

In recent years, heat pipes have been widely used in various hand held mobile electronic devices such as smart phones, tablet PCs, digital cameras. With the development of technology these devices have different user friendly features and applications; which require very high clock speeds of the processor. In general, a high clock speed generates a lot of heat, which needs to be spreaded or removed to eliminate the hot spot on the processor surface. However, it is a challenging task to achieve proper cooling of such electronic devices mentioned above because of their confined spaces and concentrated heat sources. Regarding this challenge, we introduced an ultra-thin heat pipe; this heat pipe consists of a special fiber wick structure named as "Center Fiber Wick" which can provide sufficient vapor space on the both sides of the wick structure. We also developed a cooling module that uses this kind of ultra-thin heat pipe to eliminate the hot spot issue. This cooling module consists of an ultra-thin heat pipe and a metal plate. By changing the width, the flattened thickness and the effective length of the ultra-thin heat pipe, several experiments have been conducted to characterize the thermal properties of the developed cooling module. In addition, other experiments were also conducted to determine the effects of changes in the number of heat pipes in a single module. Characterization and comparison of the module have also been conducted both experimentally and theoretically.

Fabricating cooled electronic system with liquid-cooled cold plate and thermal spreader

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

Chainer, Timothy J.; Graybill, David P.; Iyengar, Madhusudan K.

Methods are provided for facilitating cooling of an electronic component. The method includes providing a liquid-cooled cold plate and a thermal spreader associated with the cold plate. The cold plate includes multiple coolant-carrying channel sections extending within the cold plate, and a thermal conduction surface with a larger surface area than a surface area of the component to be cooled. The thermal spreader includes one or more heat pipes including multiple heat pipe sections. One or more heat pipe sections are partially aligned to a first region of the cold plate, that is, where aligned to the surface to bemore » cooled, and partially aligned to a second region of the cold plate, which is outside the first region. The one or more heat pipes facilitate distribution of heat from the electronic component to coolant-carrying channel sections of the cold plate located in the second region of the cold plate.« less

Fabricating cooled electronic system with liquid-cooled cold plate and thermal spreader

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

Chainer, Timothy J.; Graybill, David P.; Iyengar, Madhusudan K.

Methods are provided for facilitating cooling of an electronic component. The methods include providing a liquid-cooled cold plate and a thermal spreader associated with the cold plate. The cold plate includes multiple coolant-carrying channel sections extending within the cold plate, and a thermal conduction surface with a larger surface area than a surface area of the component to be cooled. The thermal spreader includes one or more heat pipes including multiple heat pipe sections. One or more heat pipe sections are partially aligned to a first region of the cold plate, that is, where aligned to the surface to bemore » cooled, and partially aligned to a second region of the cold plate, which is outside the first region. The one or more heat pipes facilitate distribution of heat from the electronic component to coolant-carrying channel sections of the cold plate located in the second region of the cold plate.« less

Cooled electronic system with liquid-cooled cold plate and thermal spreader coupled to electronic component

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

Chainer, Timothy J.; Graybill, David P.; Iyengar, Madhusudan K.

Apparatus and method are provided for facilitating cooling of an electronic component. The apparatus includes a liquid-cooled cold plate and a thermal spreader associated with the cold plate. The cold plate includes multiple coolant-carrying channel sections extending within the cold plate, and a thermal conduction surface with a larger surface area than a surface area of the component to be cooled. The thermal spreader includes one or more heat pipes including multiple heat pipe sections. One or more heat pipe sections are partially aligned to a first region of the cold plate, that is, where aligned to the surface tomore » be cooled, and partially aligned to a second region of the cold plate, which is outside the first region. The one or more heat pipes facilitate distribution of heat from the electronic component to coolant-carrying channel sections of the cold plate located in the second region of the cold plate.« less

Cooled electronic system with liquid-cooled cold plate and thermal spreader coupled to electronic component

DOEpatents

Chainer, Timothy J.; Graybill, David P.; Iyengar, Madhusudan K.; Kamath, Vinod; Kochuparambil, Bejoy J.; Schmidt, Roger R.; Steinke, Mark E.

2016-08-09

Apparatus and method are provided for facilitating cooling of an electronic component. The apparatus includes a liquid-cooled cold plate and a thermal spreader associated with the cold plate. The cold plate includes multiple coolant-carrying channel sections extending within the cold plate, and a thermal conduction surface with a larger surface area than a surface area of the component to be cooled. The thermal spreader includes one or more heat pipes including multiple heat pipe sections. One or more heat pipe sections are partially aligned to a first region of the cold plate, that is, where aligned to the surface to be cooled, and partially aligned to a second region of the cold plate, which is outside the first region. The one or more heat pipes facilitate distribution of heat from the electronic component to coolant-carrying channel sections of the cold plate located in the second region of the cold plate.

Cooled electronic system with liquid-cooled cold plate and thermal spreader coupled to electronic component

DOEpatents

Chainer, Timothy J.; Graybill, David P.; Iyengar, Madhusudan K.; Kamath, Vinod; Kochuparambil, Bejoy J.; Schmidt, Roger R.; Steinke, Mark E.

2016-04-05

Apparatus and method are provided for facilitating cooling of an electronic component. The apparatus includes a liquid-cooled cold plate and a thermal spreader associated with the cold plate. The cold plate includes multiple coolant-carrying channel sections extending within the cold plate, and a thermal conduction surface with a larger surface area than a surface area of the component to be cooled. The thermal spreader includes one or more heat pipes including multiple heat pipe sections. One or more heat pipe sections are partially aligned to a first region of the cold plate, that is, where aligned to the surface to be cooled, and partially aligned to a second region of the cold plate, which is outside the first region. The one or more heat pipes facilitate distribution of heat from the electronic component to coolant-carrying channel sections of the cold plate located in the second region of the cold plate.

Strengthening the fission reactor nuclear science and engineering program at UCLA. Final technical report

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

Okrent, D.

1997-06-23

This is the final report on DOE Award No. DE-FG03-92ER75838 A000, a three year matching grant program with Pacific Gas and Electric Company (PG and E) to support strengthening of the fission reactor nuclear science and engineering program at UCLA. The program began on September 30, 1992. The program has enabled UCLA to use its strong existing background to train students in technological problems which simultaneously are of interest to the industry and of specific interest to PG and E. The program included undergraduate scholarships, graduate traineeships and distinguished lecturers. Four topics were selected for research the first year, withmore » the benefit of active collaboration with personnel from PG and E. These topics remained the same during the second year of this program. During the third year, two topics ended with the departure o the students involved (reflux cooling in a PWR during a shutdown and erosion/corrosion of carbon steel piping). Two new topics (long-term risk and fuel relocation within the reactor vessel) were added; hence, the topics during the third year award were the following: reflux condensation and the effect of non-condensable gases; erosion/corrosion of carbon steel piping; use of artificial intelligence in severe accident diagnosis for PWRs (diagnosis of plant status during a PWR station blackout scenario); the influence on risk of organization and management quality; considerations of long term risk from the disposal of hazardous wastes; and a probabilistic treatment of fuel motion and fuel relocation within the reactor vessel during a severe core damage accident.« less

Experimental Study of Thermal Energy Storage Characteristics using Heat Pipe with Nano-Enhanced Phase Change Materials

NASA Astrophysics Data System (ADS)

Krishna, Jogi; Kishore, P. S.; Brusly Solomon, A.

2017-08-01

The paper presents experimental investigations to evaluate thermal performance of heat pipe using Nano Enhanced Phase Change Material (NEPCM) as an energy storage material (ESM) for electronic cooling applications. Water, Tricosane and nano enhanced Tricosane are used as energy storage materials, operating at different heating powers (13W, 18W and 23W) and fan speeds (3.4V and 5V) in the PCM cooling module. Three different volume percentages (0.5%, 1% and 2%) of Nano particles (Al2O3) are mixed with Tricosane which is the primary PCM. This experiment is conducted to study the temperature distributions of evaporator, condenser and PCM during the heating as well as cooling. The cooling module with heat pipe and nano enhanced Tricosane as energy storage material found to save higher fan power consumption compared to the cooling module that utilities only a heat pipe.

Heat pipe cooled heat rejection subsystem modelling for nuclear electric propulsion

NASA Astrophysics Data System (ADS)

Moriarty, Michael P.

1993-11-01

NASA LeRC is currently developing a FORTRAN based computer model of a complete nuclear electric propulsion (NEP) vehicle that can be used for piloted and cargo missions to the Moon or Mars. Proposed designs feature either a Brayton or a K-Rankine power conversion cycle to drive a turbine coupled with rotary alternators. Both ion and magnetoplasmodynamic (MPD) thrusters will be considered in the model. In support of the NEP model, Rocketdyne is developing power conversion, heat rejection, and power management and distribution (PMAD) subroutines. The subroutines will be incorporated into the NEP vehicle model which will be written by NASA LeRC. The purpose is to document the heat pipe cooled heat rejection subsystem model and its supporting subroutines. The heat pipe cooled heat rejection subsystem model is designed to provide estimate of the mass and performance of the equipment used to reject heat from Brayton and Rankine cycle power conversion systems. The subroutine models the ductwork and heat pipe cooled manifold for a gas cooled Brayton; the heat sink heat exchanger, liquid loop piping, expansion compensator, pump and manifold for a liquid loop cooled Brayton; and a shear flow condenser for a K-Rankine system. In each case, the final heat rejection is made by way of a heat pipe radiator. The radiator is sized to reject the amount of heat necessary.

Heat pipe cooled heat rejection subsystem modelling for nuclear electric propulsion

NASA Technical Reports Server (NTRS)

Moriarty, Michael P.

1993-01-01

NASA LeRC is currently developing a FORTRAN based computer model of a complete nuclear electric propulsion (NEP) vehicle that can be used for piloted and cargo missions to the Moon or Mars. Proposed designs feature either a Brayton or a K-Rankine power conversion cycle to drive a turbine coupled with rotary alternators. Both ion and magnetoplasmodynamic (MPD) thrusters will be considered in the model. In support of the NEP model, Rocketdyne is developing power conversion, heat rejection, and power management and distribution (PMAD) subroutines. The subroutines will be incorporated into the NEP vehicle model which will be written by NASA LeRC. The purpose is to document the heat pipe cooled heat rejection subsystem model and its supporting subroutines. The heat pipe cooled heat rejection subsystem model is designed to provide estimate of the mass and performance of the equipment used to reject heat from Brayton and Rankine cycle power conversion systems. The subroutine models the ductwork and heat pipe cooled manifold for a gas cooled Brayton; the heat sink heat exchanger, liquid loop piping, expansion compensator, pump and manifold for a liquid loop cooled Brayton; and a shear flow condenser for a K-Rankine system. In each case, the final heat rejection is made by way of a heat pipe radiator. The radiator is sized to reject the amount of heat necessary.

Contractor’s Meeting in Turbulence and Rotating Flows

DTIC Science & Technology

1999-08-18

pipes under turbine cooling conditions. The research results can be used for the design and fabrication of miniature heat pipes in turbine blades. The...heater used to supply the heat to the evaporator of the heat pipe was successfully fabricated . All experimental tests have been successfully completed...California, Los Angeles; D. Parekh, Georgia Tech Research Institute Rotating Miniature Heat Pipes for Turbine Blade Cooling Applications 37 Y. Cao

Energy saving potential of a two-pipe system for simultaneous heating and cooling of office buildings

DOE PAGES

Maccarini, Alessandro; Wetter, Michael; Afshari, Alireza; ...

2016-10-31

This paper analyzes the performance of a novel two-pipe system that operates one water loop to simultaneously provide space heating and cooling with a water supply temperature of around 22 °C. To analyze the energy performance of the system, a simulation-based research was conducted. The two-pipe system was modelled using the equation-based Modelica modeling language in Dymola. A typical office building model was considered as the case study. Simulations were run for two construction sets of the building envelope and two conditions related to inter-zone air flows. To calculate energy savings, a conventional four-pipe system was modelled and used formore » comparison. The conventional system presented two separated water loops for heating and cooling with supply temperatures of 45 °C and 14 °C, respectively. Simulation results showed that the two-pipe system was able to use less energy than the four-pipe system thanks to three effects: useful heat transfer from warm to cold zones, higher free cooling potential and higher efficiency of the heat pump. In particular, the two-pipe system used approximately between 12% and 18% less total annual primary energy than the four-pipe system, depending on the simulation case considered.« less

Energy saving potential of a two-pipe system for simultaneous heating and cooling of office buildings

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

Maccarini, Alessandro; Wetter, Michael; Afshari, Alireza

This paper analyzes the performance of a novel two-pipe system that operates one water loop to simultaneously provide space heating and cooling with a water supply temperature of around 22 °C. To analyze the energy performance of the system, a simulation-based research was conducted. The two-pipe system was modelled using the equation-based Modelica modeling language in Dymola. A typical office building model was considered as the case study. Simulations were run for two construction sets of the building envelope and two conditions related to inter-zone air flows. To calculate energy savings, a conventional four-pipe system was modelled and used formore » comparison. The conventional system presented two separated water loops for heating and cooling with supply temperatures of 45 °C and 14 °C, respectively. Simulation results showed that the two-pipe system was able to use less energy than the four-pipe system thanks to three effects: useful heat transfer from warm to cold zones, higher free cooling potential and higher efficiency of the heat pump. In particular, the two-pipe system used approximately between 12% and 18% less total annual primary energy than the four-pipe system, depending on the simulation case considered.« less

65. ARAII. Interior view of SL1 reactor building control piping ...

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

65. ARA-II. Interior view of SL-1 reactor building control piping for water purification system. On operating floor of building. March 21, 1958. Ineel photo no. 58-1360. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

152. ARAIII Reactor building (ARA608) Details of heater and piping ...

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

152. ARA-III Reactor building (ARA-608) Details of heater and piping pits, including instrumentation plan. Aerojet-general 880-area/GCRE-608-T-18. Date: November 1958. Ineel index code no. 063-0608-25-013-102677. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

46 CFR 182.425 - Engine exhaust cooling.

Code of Federal Regulations, 2010 CFR

2010-10-01

... otherwise provided in this paragraph, all engine exhaust pipes must be water cooled. (1) Vertical dry exhaust pipes are permissible if installed in compliance with §§ 177.405(b) and 177.970 of this chapter. (2) Horizontal dry exhaust pipes are permitted only if: (i) They do not pass through living or...

Design, Operation, and Modeling of a Vertical APCVD Reactor for Silicon Carbide Film Growth

NASA Technical Reports Server (NTRS)

DeAnna, Russell G.; Fleischman, Aaron J.; Zorman, Christian A.; Mehregany, Mehran

1998-01-01

An atmospheric pressure chemical vapor deposition (APCVD) reactor utilizing a unique vertical geometry which enables 3C-SiC films to be grown on two, 4-inch diameter Si wafers has been constructed. Contrary to expectations, 3C-SiC films grown in this reactor are thickest at the downstream end of the substrates. To better understand the reason for the thickness distribution on the wafers, an axisymmetric finite-element model of the gas flow in the reactor was constructed. The model uses the ANSYS53 Flowtran package and includes compressible and temperature-dependent fluid properties in laminar or turbulent flow. It does not include reaction chemistry or unsteady flow. The ANSYS53 results predict that the cool, inlet fluid falls through the inlet pipe and the warm, diffuser region like a jet. This jet impinges on top of the susceptor and gets diverted to the reactor side walls, where it flows to the bottom of the reactor, turns, and slowly rises along the face of the susceptor. This may explain why the SiC films are thickest at the downstream side of the wafers, as gas containing fresh reactants first passes over this region. Modeling results are presented for both one atmosphere and one half atmosphere reactor pressure.

Radiation detector system having heat pipe based cooling

DOEpatents

Iwanczyk, Jan S.; Saveliev, Valeri D.; Barkan, Shaul

2006-10-31

A radiation detector system having a heat pipe based cooling. The radiation detector system includes a radiation detector thermally coupled to a thermo electric cooler (TEC). The TEC cools down the radiation detector, whereby heat is generated by the TEC. A heat removal device dissipates the heat generated by the TEC to surrounding environment. A heat pipe has a first end thermally coupled to the TEC to receive the heat generated by the TEC, and a second end thermally coupled to the heat removal device. The heat pipe transfers the heat generated by the TEC from the first end to the second end to be removed by the heat removal device.

The Application of PVDF in Converter Cooling Pipeline

NASA Astrophysics Data System (ADS)

Geng, Man; Lu, Zhimin

2017-11-01

The structure, mechanical property, thermodynamics property, electrical aspects, radiation property and chemical property were introduced, and PVDF could satisfy the requirement of converter cooling pipe. PVDF department and pipe of distribution pipeline of converter cooling system in Debao HVDC project are used to introduce the molding process of PVDF.

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.

Heat Rejection Concepts for Lunar Fission Surface Power Applications

NASA Technical Reports Server (NTRS)

Siamidis, John

2006-01-01

This paper describes potential heat rejection design concepts for lunar surface Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for surface power applications. Surface reactors may be used for the moon to power human outposts enabling extended stays and closed loop life support. The Brayton Heat Rejection System (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Space Brayton conversion system designs tend to optimize at efficiencies of about 20 to 25 percent with radiator temperatures in the 400 K to 600 K range. A notional HRS was developed for a 100 kWe-class Brayton power system that uses a pumped water heat transport loop coupled to a water heat pipe radiator. The radiator panels employ a tube and fin construction consisting of regularly-spaced circular heat pipes contained within two composite facesheets. The water heat pipes interface to the coolant through curved sections partially contained within the cooling loop. The paper evaluates various design parameters including radiator panel orientation, coolant flow path, and facesheet thickness. Parameters were varied to compare design options on the basis of H2O pump pressure rise and required power, heat pipe unit power and radial flux, radiator area, radiator panel areal mass, and overall HRS mass.

Transient modeling of the thermohydraulic behavior of high temperature heat pipes for space reactor applications

NASA Technical Reports Server (NTRS)

Hall, Michael L.; Doster, Joseph M.

1986-01-01

Many proposed space reactor designs employ heat pipes as a means of conveying heat. Previous researchers have been concerned with steady state operation, but the transient operation is of interest in space reactor applications due to the necessity of remote startup and shutdown. A model is being developed to study the dynamic behavior of high temperature heat pipes during startup, shutdown and normal operation under space environments. Model development and preliminary results for a hypothetical design of the system are presented.

Experimental study of Siphon breaker about size effect in real scale reactor design

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

Kang, S. H.; Ahn, H. S.; Kim, J. M.

2012-07-01

Rupture accident within the pipe of a nuclear reactor is one of the main causes of a loss of coolant accident (LOCA). Siphon-breaking is a passive method that can prevent a LOCA. In this study, either a line or a hole is used as a siphon-breaker, and the effect of various parameters, such as the siphon-breaker size, pipe rupture point, pipe rupture size, and the presence of an orifice, are investigated using an experimental facility similar in size to a full-scale reactor. (authors)

Pilot-plant studies of the pipe- and pipe-cross reactors in production of granular polyphosphate fertilizers. TVA Circular Z-148. [Methods of increasing polyphosphate content of suspension fertilizers

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

Parker, B.R.; Norton, M.M.; Stumpe, T.R.

1982-01-01

Improvements have been made in the pipe-reactor or pipe-cross reactor/drum-granulator process to increase the polyphosphate content of the granular product. The goal of producing a granular APP product containing 20% of P/sub 2/O/sub 5/ as polyphosphate without adding external heat or sulfuric acid to the process has not yet been realized; however, products containing slightly more than 10% of the P/sub 2/O/sub 5/ as polyphosphate have been made without the need for external heat. Test results indicate that additions of small amounts of sulfuric acid, use of reactant NH/sub 3/:H/sub 3/PO/sub 4/ed mole ratios greater than 1.05, or use ofmore » some acid preheat may be required to consistently obtain 12% of the P/sub 2/O/sub 5/ as polyphosphate as is desired for use of the product in the preparation of suspension fertilizers. However, continued testing is being done to determine how high a mole ratio may be used successfully and to determine the effect of sulfate addition on use of the granular products for producing suspension fertilizers. The effort to obtain higher polyphosphate levels from the pipe-reactor and drum and the pipe-cross reactor and drum systems is being continued.« less

Dynamic thermal characteristics of heat pipe via segmented thermal resistance model for electric vehicle battery cooling

NASA Astrophysics Data System (ADS)

Liu, Feifei; Lan, Fengchong; Chen, Jiqing

2016-07-01

Heat pipe cooling for battery thermal management systems (BTMSs) in electric vehicles (EVs) is growing due to its advantages of high cooling efficiency, compact structure and flexible geometry. Considering the transient conduction, phase change and uncertain thermal conditions in a heat pipe, it is challenging to obtain the dynamic thermal characteristics accurately in such complex heat and mass transfer process. In this paper, a ;segmented; thermal resistance model of a heat pipe is proposed based on thermal circuit method. The equivalent conductivities of different segments, viz. the evaporator and condenser of pipe, are used to determine their own thermal parameters and conditions integrated into the thermal model of battery for a complete three-dimensional (3D) computational fluid dynamics (CFD) simulation. The proposed ;segmented; model shows more precise than the ;non-segmented; model by the comparison of simulated and experimental temperature distribution and variation of an ultra-thin micro heat pipe (UMHP) battery pack, and has less calculation error to obtain dynamic thermal behavior for exact thermal design, management and control of heat pipe BTMSs. Using the ;segmented; model, the cooling effect of the UMHP pack with different natural/forced convection and arrangements is predicted, and the results correspond well to the tests.

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

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

Ehud Greenspan

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

Three dimensional contact/impact methodology

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

Kulak, R.F.

1987-01-01

The simulation of three-dimensional interface mechanics between reactor components and structures during static contact or dynamic impact is necessary to realistically evaluate their structural integrity to off-normal loads. In our studies of postulated core energy release events, we have found that significant structure-structure interactions occur in some reactor vessel head closure designs and that fluid-structure interactions occur within the reactor vessel. Other examples in which three-dimensional interface mechanics play an important role are: (1) impact response of shipping casks containing spent fuel, (2) whipping pipe impact on reinforced concrete panels or pipe-to-pipe impact after a pipe break, (3) aircraft crashmore » on secondary containment structures, (4) missiles generated by turbine failures or tornados, and (5) drops of heavy components due to lifting accidents. The above is a partial list of reactor safety problems that require adequate treatment of interface mechanics and are discussed in this paper.« less

DEVELOPMENT OF TECHNOLOGY TO REMOTELY NAVIGATE VERTICAL PIPE ARRAYS

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

Krementz, D.; Immel, D.; Vrettos, N.

Situations exist around the Savannah River Site (SRS) and the Department of Energy (DOE) complex where it is advantageous to remotely navigate vertical pipe arrays. Specific examples are waste tanks in the SRS Tank Farms, which contain horizontal cooling coils at the tank bottom, vertical cooling coils throughout and a limited number of access points or ''risers''. These factors limit accessibility to many parts of these tanks by conventional means. Pipe Traveler technology has been developed to address these issues. The Pipe Traveler addresses these issues by using the vertical cooling coils as its medium of travel. The unit operatesmore » by grabbing a pipe using dual grippers located on either side of the equipment. Once securely attached to the pipe a drive wheel is extended to come in contact with the pipe. Rotation of the drive wheel causes the unit to rotate around the pipe. This action is continued until the second set of grippers is aligned with the next pipe. Extension pistons are actuated to extend the second set of grippers in contact with a second pipe. The second set of grippers is then actuated to grasp the pipe. The first set of grippers releases the original pipe and the process is repeated until the unit reaches its desired location. Once at the tool deployment location the desired tool may be used. The current design has proven the concept of pipe-to-pipe navigation. Testing of the Pipe Traveler has proven its ability to transfer itself from one pipe to another.« less

Passive cryogenic cooling of electrooptics with a heat pipe/radiator.

PubMed

Nelson, B E; Goldstein, G A

1974-09-01

The current status of the heat pipe is discussed with particular emphasis on applications to cryogenic thermal control. The competitive nature of the passive heat pipe/radiator system is demonstrated through a comparative study with other candidate systems for a 1-yr mission. The mission involves cooling a spaceborne experiment to 100 K while it dissipates 10 W.

System for Cooling of Electronic Components

NASA Astrophysics Data System (ADS)

Vasil'ev, L. L.; Grakovich, L. P.; Dragun, L. A.; Zhuravlev, A. S.; Olekhnovich, V. A.; Rabetskii, M. I.

2017-01-01

Results of computational and experimental investigations of heat pipes having a predetermined thermal resistance and a system based on these pipes for air cooling of electronic components and diode assemblies of lasers are presented. An efficient compact cooling system comprising heat pipes with an evaporator having a capillary coating of a caked copper powder and a condenser having a developed outer finning, has been deviced. This system makes it possible to remove, to the ambient air, a heat flow of power more than 300 W at a temperature of 40-50°C.

Analysis, Verification, and Application of Equations and Procedures for Design of Exhaust-pipe Shrouds

NASA Technical Reports Server (NTRS)

Ellerbrock, Herman H.; Wcislo, Chester R.; Dexter, Howard E.

1947-01-01

Investigations were made to develop a simplified method for designing exhaust-pipe shrouds to provide desired or maximum cooling of exhaust installations. Analysis of heat exchange and pressure drop of an adequate exhaust-pipe shroud system requires equations for predicting design temperatures and pressure drop on cooling air side of system. Present experiments derive such equations for usual straight annular exhaust-pipe shroud systems for both parallel flow and counter flow. Equations and methods presented are believed to be applicable under certain conditions to the design of shrouds for tail pipes of jet engines.

Qualification of coolants and cooling pipes for future high-energy-particle detectors

NASA Astrophysics Data System (ADS)

Ilie, Sorin; Tavlet, Marc

2001-12-01

In the next generation of high-energy-particle detectors to be installed at the Large Hadron Collider (LHC) at CERN, materials and components will be exposed to a significant level of ionising radiation. Silicon detectors and related electronics will have to be cooled down to -20 °C and therefore appropriate cooling fluids and cooling pipes have to be selected. Analytical methods such as UV-visible and FT-IR spectrometries, electronic microscopy and gas chromatography were used to characterise the radiation-induced effects on some organic coolants irradiated with both gamma and neutron fields. Some impurities were identified as a major source for radio-induced polymerisation and also for hydrofluoric acid (HF) evolution. Mechanical tests were performed to assess the operability of the rubber hoses and plastic pipes. Possible synergistic effects between the pipe material and the environment had to be considered.

A thermosyphon heat pipe cooler for high power LEDs cooling

NASA Astrophysics Data System (ADS)

Li, Ji; Tian, Wenkai; Lv, Lucang

2016-08-01

Light emitting diode (LED) cooling is facing the challenge of high heat flux more seriously with the increase of input power and diode density. The proposed unique thermosyphon heat pipe heat sink is particularly suitable for cooling of high power density LED chips and other electronics, which has a heat dissipation potential of up to 280 W within an area of 20 mm × 22 mm (>60 W/cm2) under natural air convection. Meanwhile, a thorough visualization investigation was carried out to explore the two phase flow characteristics in the proposed thermosyphon heat pipe. Implementing this novel thermosyphon heat pipe heat sink in the cooling of a commercial 100 W LED integrated chip, a very low apparent thermal resistance of 0.34 K/W was obtained under natural air convection with the aid of the enhanced boiling heat transfer at the evaporation side and the enhanced natural air convection at the condensation side.

Passive environmental temperature control system

DOEpatents

Corliss, John M.; Stickford, George H.

1981-01-01

Passive environmental heating and cooling systems are described, which utilize heat pipes to transmit heat to or from a thermal reservoir. In a solar heating system, a heat pipe is utilized to carry heat from a solar heat absorber plate that receives sunlight, through a thermal insulation barrier, to a heat storage wall, with the outer end of the pipe which is in contact with the solar absorber being lower than the inner end. The inclining of the heat pipe assures that the portion of working fluid, such as Freon, which is in a liquid phase will fall by gravity to the outer end of the pipe, thereby assuring diode action that prevents the reverse transfer of heat from the reservoir to the outside on cool nights. In a cooling system, the outer end of the pipe which connects to a heat dissipator, is higher than the inner end that is coupled to a cold reservoir, to allow heat transfer only out of the reservoir to the heat dissipator, and not in the reverse direction.

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

Feng Xie; Hong Li; Jianzhu Cao

A reform will be implemented in the helium purification system of the 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) in China. The measurement of the gamma dose rates of facilities, including valves, pipes, dust filter, etc., in the purification system of the HTR-10, has been performed. The results indicated that most radiation nuclides are concentrated in the dust filter and facilities at the entrance of the helium purification system upstream of the dust filter. Other facilities have the same gamma dose rate level as the background. Based on the previous study and experiences in AVR, the measurement results canmore » be understood that the radioactive dust carried by the helium gas was filtered by the dust filter. It provides important insights for the decontamination and decommissioning of facilities in the primary loop, especially in the helium purification system of the HTR-10 as well as the High Temperature Reactor-Pebble bed Modules (HTR-PM). (authors)« less

Design consideration for a nuclear electric propulsion system

NASA Technical Reports Server (NTRS)

Phillips, W. M.; Pawlik, E. V.

1978-01-01

A study is currently underway to design a nuclear electric propulsion vehicle capable of performing detailed exploration of the outer-planets. Primary emphasis is on the power subsystem. Secondary emphasis includes integration into a spacecraft, and integration with the thrust subsystem and science package or payload. The results of several design iterations indicate an all-heat-pipe system offers greater reliability, elimination of many technology development areas and a specific weight of under 20 kg/kWe at the 400 kWe power level. The system is compatible with a single Shuttle launch and provides greater safety than could be obtained with designs using pumped liquid metal cooling. Two configurations, one with the reactor and power conversion forward on the spacecraft with the ion engines aft and the other with reactor, power conversion and ion engines aft were selected as dual baseline designs based on minimum weight, minimum required technology development and maximum growth potential and flexibility.

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.

Testing of a Methane Cryogenic Heat Pipe with a Liquid Trap Turn-Off Feature for use on Space Interferometer Mission (SIM)

NASA Technical Reports Server (NTRS)

Cepeda-Rizo, Juan; Krylo, Robert; Fisher, Melanie; Bugby, David C.

2011-01-01

Camera cooling for SIM presents three thermal control challenges; stable operation at 163K (110 C), decontamination heating to +20 C, and a long span from the cameras to the radiator. A novel cryogenic cooling system based on a methane heat pipe meets these challenges. The SIM thermal team, with the help of heat pipe vendor ATK, designed and tested a complete, low temperature, cooling system. The system accommodates the two SIM cameras with a double-ended conduction bar, a single methane heat pipe, independent turn-off devices, and a flight-like radiator. The turn ]off devices consist of a liquid trap, for removing the methane from the pipe, and an electrical heater to raise the methane temperature above the critical point thus preventing two-phase operation. This is the first time a cryogenic heat pipe has been tested at JPL and is also the first heat pipe to incorporate the turn-off features. Operation at 163K with a methane heat pipe is an important new thermal control capability for the lab. In addition, the two turn-off technologies enhance the "bag of tricks" available to the JPL thermal community. The successful test program brings this heat pipe to a high level of technology readiness.

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.

Silicon Heat Pipe Array

NASA Technical Reports Server (NTRS)

Yee, Karl Y.; Ganapathi, Gani B.; Sunada, Eric T.; Bae, Youngsam; Miller, Jennifer R.; Beinsford, Daniel F.

2013-01-01

Improved methods of heat dissipation are required for modern, high-power density electronic systems. As increased functionality is progressively compacted into decreasing volumes, this need will be exacerbated. High-performance chip power is predicted to increase monotonically and rapidly with time. Systems utilizing these chips are currently reliant upon decades of old cooling technology. Heat pipes offer a solution to this problem. Heat pipes are passive, self-contained, two-phase heat dissipation devices. Heat conducted into the device through a wick structure converts the working fluid into a vapor, which then releases the heat via condensation after being transported away from the heat source. Heat pipes have high thermal conductivities, are inexpensive, and have been utilized in previous space missions. However, the cylindrical geometry of commercial heat pipes is a poor fit to the planar geometries of microelectronic assemblies, the copper that commercial heat pipes are typically constructed of is a poor CTE (coefficient of thermal expansion) match to the semiconductor die utilized in these assemblies, and the functionality and reliability of heat pipes in general is strongly dependent on the orientation of the assembly with respect to the gravity vector. What is needed is a planar, semiconductor-based heat pipe array that can be used for cooling of generic MCM (multichip module) assemblies that can also function in all orientations. Such a structure would not only have applications in the cooling of space electronics, but would have commercial applications as well (e.g. cooling of microprocessors and high-power laser diodes). This technology is an improvement over existing heat pipe designs due to the finer porosity of the wick, which enhances capillary pumping pressure, resulting in greater effective thermal conductivity and performance in any orientation with respect to the gravity vector. In addition, it is constructed of silicon, and thus is better suited for the cooling of semiconductor devices.

Double Retort System for Materials Compatibility Testing

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

V. Munne; EV Carelli

2006-02-23

With Naval Reactors (NR) approval of the Naval Reactors Prime Contractor Team (NRPCT) recommendation to develop a gas cooled reactor directly coupled to a Brayton power conversion system as the Space Nuclear Power Plant (SNPP) for Project Prometheus (References a and b) there was a need to investigate compatibility between the various materials to be used throughout the SNPP. Of particular interest was the transport of interstitial impurities from the nickel-base superalloys, which were leading candidates for most of the piping and turbine components to the refractory metal alloys planned for use in the reactor core. This kind of contaminationmore » has the potential to affect the lifetime of the core materials. This letter provides technical information regarding the assembly and operation of a double retort materials compatibility testing system and initial experimental results. The use of a double retort system to test materials compatibility through the transfer of impurities from a source to a sink material is described here. The system has independent temperature control for both materials and is far less complex than closed loops. The system is described in detail and the results of three experiments are presented.« less

Development and optimization of water treatment reactors using TiO2-modified polymer beads with a refractive index identical to that of water

NASA Astrophysics Data System (ADS)

Myoga, Arata; Iwashita, Ryutaro; Unno, Noriyuki; Satake, Shin-ichi; Taniguchi, Jun; Yuki, Kazuhisa; Seki, Yohji

2018-03-01

Various water purification reactors were constructed using beads of TiO2-coated MEXFLON, which is a fluoropolymer exhibiting a refractive index identical to that of water. The performance of these reactors was evaluated in a recirculation experiment utilizing an aqueous solution of methylene blue. Reactor pipes (length = 150 mm, internal diameter = 10 mm) were made of a fluorinated ethylene polymer with a refractive index of 1.338 and contained 206-bead clusters. A UV lamp was used to irradiate eight reactor pipes surrounding it. The above-mentioned eight bead-packed pipes were connected both in series and in parallel, and the performances of these two reactor types were compared. A pseudo-first-order rate constant of 0.70 h- 1 was obtained for the series connection, whereas the corresponding value for the parallel connection was 1.5 times smaller, confirming the effectiveness of increasing the reaction surface by employing a larger number of beads.

Development and optimization of water treatment reactors using TiO2-modified polymer beads with a refractive index identical to that of water

NASA Astrophysics Data System (ADS)

Myoga, Arata; Iwashita, Ryutaro; Unno, Noriyuki; Satake, Shin-ichi; Taniguchi, Jun; Yuki, Kazuhisa; Seki, Yohji

2018-06-01

Various water purification reactors were constructed using beads of TiO2-coated MEXFLON, which is a fluoropolymer exhibiting a refractive index identical to that of water. The performance of these reactors was evaluated in a recirculation experiment utilizing an aqueous solution of methylene blue. Reactor pipes (length = 150 mm, internal diameter = 10 mm) were made of a fluorinated ethylene polymer with a refractive index of 1.338 and contained 206-bead clusters. A UV lamp was used to irradiate eight reactor pipes surrounding it. The above-mentioned eight bead-packed pipes were connected both in series and in parallel, and the performances of these two reactor types were compared. A pseudo-first-order rate constant of 0.70 h- 1 was obtained for the series connection, whereas the corresponding value for the parallel connection was 1.5 times smaller, confirming the effectiveness of increasing the reaction surface by employing a larger number of beads.

ETR BUILDING, TRA642, INTERIOR. BASEMENT. LIQUID SODIUM PIPING INSIDE CUBICLE ...

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

ETR BUILDING, TRA-642, INTERIOR. BASEMENT. LIQUID SODIUM PIPING INSIDE CUBICLE SHOWN IN ID-33-G-101. INL NEGATIVE NO. HD24-3-4. Mike Crane, Photographer, 11/2000 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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.

Heat-pipe planets

NASA Astrophysics Data System (ADS)

Moore, William B.; Simon, Justin I.; Webb, A. Alexander G.

2017-09-01

Observations of the surfaces of all terrestrial bodies other than Earth reveal remarkable but unexplained similarities: endogenic resurfacing is dominated by plains-forming volcanism with few identifiable centers, magma compositions are highly magnesian (mafic to ultra-mafic), tectonic structures are dominantly contractional, and ancient topographic and gravity anomalies are preserved to the present. Here we show that cooling via volcanic heat pipes may explain these observations and provide a universal model of the way terrestrial bodies transition from a magma-ocean state into subsequent single-plate, stagnant-lid convection or plate tectonic phases. In the heat-pipe cooling mode, magma moves from a high melt-fraction asthenosphere through the lithosphere to erupt and cool at the surface via narrow channels. Despite high surface heat flow, the rapid volcanic resurfacing produces a thick, cold, and strong lithosphere which undergoes contractional strain forced by downward advection of the surface toward smaller radii. We hypothesize that heat-pipe cooling is the last significant endogenic resurfacing process experienced by most terrestrial bodies in the solar system, because subsequent stagnant-lid convection produces only weak tectonic deformation. Terrestrial exoplanets appreciably larger than Earth may remain in heat-pipe mode for much of the lifespan of a Sun-like star.

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

Ballinger, Ronald G

A material that resists lead-bismuth eutectic (LBE) attack and retains its strength at 700°C would be an enabling technology for LBE-cooled reactors. No single alloy currently exists that can economically meet the required performance criteria of high strength and corrosion resistance. A Functionally Graded Composite (FGC) was developed with layers engineered to perform these functions. F91 was chosen as the structural layer of the composite for its strength and radiation resistance. Fe-12Cr-2Si, an alloy developed from previous work in the Fe-Cr-Si system, was chosen as the corrosion-resistant cladding layer because of its chemical similarity to F91 and its superior corrosionmore » resistance in both oxidizing and reducing environments. Fe-12Cr-2Si experienced minimal corrosion due to its self-passivation in oxidizing and reducing environments. Extrapolated corrosion rates are below one micron per year at 700ï°C. Corrosion of F91 was faster, but predictable and manageable. Diffusion studies showed that 17 microns of the cladding layer will be diffusionally diluted during the three year life of fuel cladding. 33 microns must be accounted for during the sixty year life of coolant piping. 5 cm coolant piping and 6.35 mm fuel cladding preforms were produced on a commercial scale by weld-overlaying Fe-12Cr-2Si onto F91 billets and co-extruding them. An ASME certified weld was performed followed by the prescribed quench-and-tempering heat treatment for F91. A minimal heat affected zone was observed, demonstrating field weldability. Finally, corrosion tests were performed on the fabricated FGC at 700ï°C after completely breaching the cladding in a small area to induce galvanic corrosion at the interface. None was observed. This FGC has significant impacts on LBE reactor design. The increases in outlet temperature and coolant velocity allow a large increase in power density, leading to either a smaller core for the same power rating or more power output for the same size core. This FGC represents an enabling technology for LBE cooled fast reactors.« less

Apparatus for storing hydrogen isotopes

DOEpatents

McMullen, John W.; Wheeler, Michael G.; Cullingford, Hatice S.; Sherman, Robert H.

1985-01-01

An improved method and apparatus for storing isotopes of hydrogen (especially tritium) are provided. The hydrogen gas(es) is (are) stored as hydrides of material (for example uranium) within boreholes in a block of copper. The mass of the block is critically important to the operation, as is the selection of copper, because no cooling pipes are used. Because no cooling pipes are used, there can be no failure due to cooling pipes. And because copper is used instead of stainless steel, a significantly higher temperature can be reached before the eutectic formation of uranium with copper occurs, (the eutectic of uranium with the iron in stainless steel forming at a significantly lower temperature).

Modular Cooling Components

NASA Technical Reports Server (NTRS)

Eastman, G. Yale; Dussinger, Peter M.; Hartenstine, John R.

1994-01-01

Three modular heat-transfer components designed for use together or separately. Simple mechanical connections facilitate assembly of these and related heat-transfer components into cooling systems of various configurations, such as to cool laboratory equipment rearranged for different experiments. Components are clamp-on cold plate, cold plate attached to flexible heat pipe, and thermal-bus receptacle. Clamp-on cold plate moved to any convenient location for attachment of equipment cooled by it, then clamped onto thermal bus. Heat from equipment conducted through plate and into coolant. Thermal-bus receptacle integral with thermal bus. Includes part of thermal bus to which clamp-on cold plate attached, plus tapered socket into which condenser end of flexible heat pipe plugged. Thermal-bus receptacle includes heat-pipe wick structure using coolant in bus to enhance transfer of heat from cold plate.

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.

An electronic cryoprobe for cryosurgery using heat pipes and thermoelectric coolers: a preliminary report.

PubMed

Hamilton, A; Hu, J

1993-01-01

A hand-held fully electrically powered and programmable cryoprobe for general-purpose cryosurgery and cryotherapy has been developed. By combining the technologies of thermoelectric cooling and heat pipes, the temperature at the tip of the probe can easily reach -50 to -60 degrees C. It can hold below -40 degrees C when it cools a load of 10 W at the tip. Previous efforts developing cryoprobes made of thermoelectric modules have been hindered by the inherent characteristics of commercially available thermoelectric coolers: low efficiency, size and inflexible shape and very sensitive to heat intensity and thermal insulation. Matching thermoelectrics with heat pipes uses the advantages of both technologies. In the cryoprobe the heat pipe is used to focus and transport the cooling power of multi-thermoelectric modules. The heat flux for the thermoelectric modules is reduced and their efficiencies are increased. The transport of heat by a heat pipe also allows flexible access to treated spots of patients.

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.

DEVELOPMENT OF A CHARGING/COLLECTING DEVICE FOR HIGH RESISTIVITY DUST USING COOLED ELECTRODES

EPA Science Inventory

The paper discusses a charging/collecting device for high-resistivity fly ash, developed to control back-ionization by cooling the collector electrode internally with water. The device consists of parallel 6.0 cm pipes with corona wires suspended between them. The pipes provide a...

Fabrication of lithium/C-103 alloy heat pipes for sharp leading edge cooling

NASA Astrophysics Data System (ADS)

Ai, Bangcheng; Chen, Siyuan; Yu, Jijun; Lu, Qin; Han, Hantao; Hu, Longfei

2018-05-01

In this study, lithium/C-103 alloys heat pipes are proposed for sharp leading edge cooling. Three models of lithium/C-103 alloy heat pipes were fabricated. And their startup properties were tested by radiant heat tests and aerothermal tests. It is found that the startup temperature of lithium heat pipe was about 860 °C. At 1000 °C radiant heat tests, the operating temperature of lithium/C-103 alloy heat pipe is lower than 860 °C. Thus, startup failure occurs. At 1100 °C radiant heat tests and aerothermal tests, the operating temperature of lithium/C-103 alloy heat pipe is higher than 860 °C, and the heat pipe starts up successfully. The startup of lithium/C-103 alloy heat pipe decreases the leading edge temperature effectively, which endows itself good ablation resistance. After radiant heat tests and aerothermal tests, all the heat pipe models are severely oxidized because of the C-103 poor oxidation resistance. Therefore, protective coatings are required for further applications of lithium/C-103 alloy heat pipes.

Rust Inhibitor And Fungicide For Cooling Systems

NASA Technical Reports Server (NTRS)

Adams, James F.; Greer, D. Clay

1988-01-01

Mixture of benzotriazole, benzoic acid, and fungicide prevents growth of rust and fungus. Water-based cooling mixture made from readily available materials prevents formation of metallic oxides and growth of fungi in metallic pipes. Coolant remains clear and does not develop thick sludge tending to collect in low points in cooling systems with many commercial rust inhibitors. Coolant compatible with iron, copper, aluminum, and stainless steel. Cannot be used with cadmium or cadmium-plated pipes.

Probabilistic pipe fracture evaluations for leak-rate-detection applications

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

Rahman, S.; Ghadiali, N.; Paul, D.

1995-04-01

Regulatory Guide 1.45, {open_quotes}Reactor Coolant Pressure Boundary Leakage Detection Systems,{close_quotes} was published by the U.S. Nuclear Regulatory Commission (NRC) in May 1973, and provides guidance on leak detection methods and system requirements for Light Water Reactors. Additionally, leak detection limits are specified in plant Technical Specifications and are different for Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). These leak detection limits are also used in leak-before-break evaluations performed in accordance with Draft Standard Review Plan, Section 3.6.3, {open_quotes}Leak Before Break Evaluation Procedures{close_quotes} where a margin of 10 on the leak detection limit is used in determining the crackmore » size considered in subsequent fracture analyses. This study was requested by the NRC to: (1) evaluate the conditional failure probability for BWR and PWR piping for pipes that were leaking at the allowable leak detection limit, and (2) evaluate the margin of 10 to determine if it was unnecessarily large. A probabilistic approach was undertaken to conduct fracture evaluations of circumferentially cracked pipes for leak-rate-detection applications. Sixteen nuclear piping systems in BWR and PWR plants were analyzed to evaluate conditional failure probability and effects of crack-morphology variability on the current margins used in leak rate detection for leak-before-break.« less

Cryostat system for investigation on new neutron moderator materials at reactor TRIGA PUSPATI

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

Dris, Zakaria bin, E-mail: zakariadris@gmail.com; Centre for Nuclear Energy, Universiti Tenaga Nasional; Mohamed, Abdul Aziz bin

2016-01-22

A simple continuous flow (SCF) cryostat was designed to investigate the neutron moderation of alumina in high temperature co-ceramic (HTCC) and polymeric materials such as Teflon under TRIGA neutron environment using a reflected neutron beam from a monochromator. Cooling of the cryostat will be carried out using liquid nitrogen. The cryostat will be built with an aluminum holder for moderator within stainless steel cylinder pipe. A copper thermocouple will be used as the temperature sensor to monitor the moderator temperature inside the cryostat holder. Initial measurements of neutron spectrum after neutron passing through the moderating materials have been carried outmore » using a neutron spectrometer.« less

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.

Thermal performance analysis of a flat heat pipe working with carbon nanotube-water nanofluid for cooling of a high heat flux heater

NASA Astrophysics Data System (ADS)

Arya, A.; Sarafraz, M. M.; Shahmiri, S.; Madani, S. A. H.; Nikkhah, V.; Nakhjavani, S. M.

2018-04-01

Experimental investigation on the thermal performance of a flat heat pipe working with carbon nanotube nanofluid is conducted. It is used for cooling a heater working at high heat flux conditions up to 190 kW/m2. The heat pipe is fabricated from aluminium and is equipped with rectangular fin for efficient cooling of condenser section. Inside the heat pipe, a screen mesh was inserted as a wick structure to facilitate the capillary action of working fluid. Influence of different operating parameters such as heat flux, mass concentration of carbon nanotubes and filling ratio of working fluid on thermal performance of heat pipe and its thermal resistance are investigated. Results showed that with an increase in heat flux, the heat transfer coefficient in evaporator section of the heat pipe increases. For filling ratio, however, there is an optimum value, which was 0.8 for the test heat pipe. In addition, CNT/water enhanced the heat transfer coefficient up to 40% over the deionized water. Carbon nanotubes intensified the thermal performance of wick structure by creating a fouling layer on screen mesh structure, which changes the contact angle of liquid with the surface, intensifying the capillary forces.

Heat pipe cooling system with sensible heat sink

NASA Technical Reports Server (NTRS)

Silverstein, Calvin C.

1988-01-01

A heat pipe cooling system which employs a sensible heat sink is discussed. With this type of system, incident aerodynamic heat is transported via a heat pipe from the stagnation region to the heat sink and absorbed by raising the temperature of the heat sink material. The use of a sensible heat sink can be advantageous for situations where the total mission heat load is limited, as it is during re-entry, and a suitable radiation sink is not available.

Cooling system for a nuclear reactor

DOEpatents

Amtmann, Hans H.

1982-01-01

A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

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.

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.

Research and Development of Ultra-High Strength X100 Welded Pipe

NASA Astrophysics Data System (ADS)

Chuanguo, Zhang; Lei, Zheng; Ping, Hu; Bei, Zhang; Kougen, Wu; Weifeng, Huang

Ultra-high strength X100 welded pipe can be used in the construction of long distance oil and gas pipeline to improve transmission capacity and reduce operation cost. By using the way of thermo-simulation and pilot rolling, the CCT (Continuous Cooling Transformation) diagram and the relationship between ACC (Accelerated Cooling) parameters, microstructure and mechanical properties were studied for the designed X100 pipeline steel with low carbon, high manganese and niobium micro-alloyed composition in lab. The analysis of CCT diagram indicates that the suitable hardness and microstructure can be obtained in the cooling rate of 20 80°C/sec. The pilot rolling results show that the ACC cooling start temperature below Ar3 phase transformation point is beneficial to increase uniform elongation, and the cooling stop temperature of 150 350°C is helpful to obtain high strength and toughness combination. Based on the research conclusions, the X100 plate and UOE pipe with dimension in O.D.1219×W.T.14.8mm, O.D.1219×W.T.17.8mm, designed for the natural gas transmission pipeline, were trial produced. The manufactured pipe body impact absorbed energy at -10°C is over 250J. The DWTT shear area ratio at 0°C is over 85%. The transverse strength meets the X100 grade requirement, and uniform elongation is over 4%. The X100 plate and UOE pipe with dimension in O.D.711×W.T.20.0mm, O.D.711×W.T.12.5mm, designed for an offshore engineering, were also trial produced. The average impact absorbed energy of pipe body at -30°C is over 200J. The average impact absorbed energy of HAZ (Heat-affected zone) and WM (Welded Seam) at -30°C is over 100J. And the good pipe shapes were obtained

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)

An underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, V.E.

1988-05-17

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, Viktor E.

1989-01-01

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

Wind tunnel data of the analysis of heat pipe and wind catcher technology for the built environment.

PubMed

Calautit, John Kaiser; Chaudhry, Hassam Nasarullah; Hughes, Ben Richard

2015-12-01

The data presented in this article were the basis for the study reported in the research articles entitled 'Climate responsive behaviour heat pipe technology for enhanced passive airside cooling' by Chaudhry and Hughes [10] which presents the passive airside cooling capability of heat pipes in response to gradually varying external temperatures and related to the research article "CFD and wind tunnel study of the performance of a uni-directional wind catcher with heat transfer devices" by Calautit and Hughes [1] which compares the ventilation performance of a standard roof mounted wind catcher and wind catcher incorporating the heat pipe technology. Here, we detail the wind tunnel test set-up and inflow conditions and the methodologies for the transient heat pipe experiment and analysis of the integration of heat pipes within the control domain of a wind catcher design.

Effects of phosphate addition on biofilm bacterial communities and water quality in annular reactors equipped with stainless steel and ductile cast iron pipes.

PubMed

Jang, Hyun-Jung; Choi, Young-June; Ro, Hee-Myong; Ka, Jong-Ok

2012-02-01

The impact of orthophosphate addition on biofilm formation and water quality was studied in corrosion-resistant stainless steel (STS) pipe and corrosion-susceptible ductile cast iron (DCI) pipe using cultivation and culture-independent approaches. Sample coupons of DCI pipe and STS pipe were installed in annular reactors, which were operated for 9 months under hydraulic conditions similar to a domestic plumbing system. Addition of 5 mg/L of phosphate to the plumbing systems, under low residual chlorine conditions, promoted a more significant growth of biofilm and led to a greater rate reduction of disinfection by-products in DCI pipe than in STS pipe. While the level of THMs (trihalomethanes) increased under conditions of low biofilm concentration, the levels of HAAs (halo acetic acids) and CH (chloral hydrate) decreased in all cases in proportion to the amount of biofilm. It was also observed that chloroform, the main species of THM, was not readily decomposed biologically and decomposition was not proportional to the biofilm concentration; however, it was easily biodegraded after the addition of phosphate. Analysis of the 16S rDNA sequences of 102 biofilm isolates revealed that Proteobacteria (50%) was the most frequently detected phylum, followed by Firmicutes (10%) and Actinobacteria (2%), with 37% of the bacteria unclassified. Bradyrhizobium was the dominant genus on corroded DCI pipe, while Sphingomonas was predominant on non-corroded STS pipe. Methylobacterium and Afipia were detected only in the reactor without added phosphate. PCR-DGGE analysis showed that the diversity of species in biofilm tended to increase when phosphate was added regardless of the pipe material, indicating that phosphate addition upset the biological stability in the plumbing systems.

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.

Radiation Shielding Design and Orientation Considerations for a 1 kWe Heat Pipe Cooled Reactor Utilized to Bore Through the Ice Caps of Mars

NASA Astrophysics Data System (ADS)

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

2006-01-01

The goal in designing any space power system is to develop a system able to meet the mission requirements for success while minimizing the overall costs. The mission requirements for the this study was to develop a reactor (with Stirling engine power conversion) and shielding configuration able to fit, along with all the other necessary science equipment, in a Cryobot 3 m high with ~0.5 m diameter hull, produce 1 kWe for 5yrs, and not adversely affect the mission science by keeping the total integrated dose to the science equipment below 150 krad. Since in most space power missions the overall system mass dictates the mission cost, the shielding designs in this study incorporated Martian water extracted at the startup site in order to minimize the tungsten and LiH mass loading at launch. Different reliability and mass minimization concerns led to three design configuration evolutions. With the help of implementing Martian water and configuring the reactor as far from the science equipment as possible, the needed tungsten and LiH shield mass was minimized. This study further characterizes the startup dose and the necessary mission requirements in order to ensure integrity of the surface equipment during reactor startup phase.

Design study of a 120-keV, He-3 neutral beam injector

NASA Astrophysics Data System (ADS)

Blum, A. S.; Barr, W. L.; Dexter, W. L.; Moir, R. W.; Wilcox, T. P.; Fink, J. H.

1981-01-01

A design for a 120-keV, 2.3-MW, He-3 neutral beam injector for use on a D-(He-3) fusion reactor is described. The constraint that limits operating life when injecting He is its high sputtering rate. The sputtering is partly controlled by using an extra grid to prevent ion flow from the neutralizer duct to the electron suppressor grid, but a tradeoff between beam current and operating life is still required. Hollow grid wires functioning as mercury heat pipes cool the grid and enable steady state operation. Voltage holding and radiation effects on the acceleration grid structure are discussed. The vacuum system is also briefly described, and the use of a direct energy converter to recapture energy from unneutralized ions exiting the neutralizer is also analyzed. Of crucial importance to the technical feasibility of the (He-3)-burning reactor are the injector efficiency and cost; these are 53% and $5.5 million, respectively, when power supplies are included.

Design study of a 120-keV,3He neutral beam injector

NASA Astrophysics Data System (ADS)

Blum, A. S.; Barr, W. L.; Dexter, W. L.; Fink, J. H.; Moir, R. W.; Wilcox, T. P.

1981-01-01

We describe a design for a 120-keV, 2.3-MW,3He neutral beam injector for use on a D-3He fusion reactor. The constraint that limits operating life when injecting He is its high sputtering rate. The sputtering is partly controlled by using an extra grid to prevent ion flow from the neutralizer duct to the electron suppressor grid, but a tradeoff between beam current and operating life is still required. Hollow grid wires functioning as mercury heat pipes cool the grid and enable steady state operation. Voltage holding and radiation effects on the acceleration grid structure are discussed. We also briefly describe the vacuum system and analyze use of a direct energy converter to recapture energy from unneutralized ions exiting the neutralizer. Of crucial importance to the technical feasibility of the3He-burning reactor are the injector efficiency and cost; these are 53% and 5.5 million, respectively, when power supplies are included.

Method and apparatus for storing hydrogen isotopes. [stored as uranium hydride in a block of copper

DOEpatents

McMullen, J.W.; Wheeler, M.G.; Cullingford, H.S.; Sherman, R.H.

1982-08-10

An improved method and apparatus for storing isotopes of hydrogen (especially tritium) are provided. The hydrogen gas is stored as hydrides of material (for example uranium) within boreholes in a block of copper. The mass of the block is critically important to the operation, as is the selection of copper, because no cooling pipes are used. Because no cooling pipes are used, there can be no failure due to cooling pipes. And because copper is used instead of stainless steel, a significantly higher temperature can be reached before the eutectic formation of uranium with copper occurs, (the eutectic of uranium with the iron in stainless steel forms at a significantly lower temperature).

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.

Air Conditioner/Dehumidifier

NASA Technical Reports Server (NTRS)

1986-01-01

An ordinary air conditioner in a very humid environment must overcool the room air, then reheat it. Mr. Dinh, a former STAC associate, devised a heat pipe based humidifier under a NASA Contract. The system used heat pipes to precool the air; the air conditioner's cooling coil removes heat and humidity, then the heat pipes restore the overcooled air to a comfortable temperature. The heat pipes use no energy, and typical savings are from 15-20%. The Dinh Company also manufactures a "Z" coil, a retrofit cooling coil which may be installed on an existing heater/air conditioner. It will also provide free hot water. The company has also developed a photovoltaic air conditioner and solar powered water pump.

Status of the Development of Low Cost Radiator for Surface Fission Power - II

NASA Technical Reports Server (NTRS)

Tarau, Calin; Maxwell, Taylor; Anderson, William G.; Wagner, Corey; Wrosch, Matthew; Briggs, Maxwell H.

2016-01-01

NASA Glenn Research Center (GRC) is developing fission power system technology for future Lunar and Martian surface power applications. The systems are envisioned in the 10 to 100kWe range and have an anticipated design life of 8 to 15 years with no maintenance. NASA GRC is currently setting up a 55 kWe non-nuclear system ground test in thermal-vacuum to validate technologies required to transfer reactor heat, convert the heat into electricity, reject waste heat, process the electrical output, and demonstrate overall system performance. The paper reports on the development of the heat pipe radiator to reject the waste heat from the Stirling convertors. Reducing the radiator mass, size, and cost is essential to the success of the program. To meet these goals, Advanced Cooling Technologies, Inc. (ACT) and Vanguard Space Technologies, Inc. (VST) are developing a single facesheet radiator with heat pipes directly bonded to the facesheet. The facesheet material is a graphite fiber reinforced composite (GFRC) and the heat pipes are titanium/water Variable Conductance Heat Pipes (VCHPs). By directly bonding a single facesheet to the heat pipes, several heavy and expensive components can be eliminated from the traditional radiator design such as, POCO"TM" foam saddles, aluminum honeycomb, and a second facesheet. As mentioned in previous papers by the authors, the final design of the waste heat radiator is described as being modular with independent GFRC panels for each heat pipe. The present paper reports on test results for a single radiator module as well as a radiator cluster consisting of eight integral modules. These tests were carried out in both ambient and vacuum conditions. While the vacuum testing of the single radiator module was performed in the ACT's vacuum chamber, the vacuum testing of the eight heat pipe radiator cluster took place in NASA GRC's vacuum chamber to accommodate the larger size of the cluster. The results for both articles show good agreement with the predictions and are presented in the paper.

Effect of pipe corrosion scales on chlorine dioxide consumption in drinking water distribution systems.

PubMed

Zhang, Zhe; Stout, Janet E; Yu, Victor L; Vidic, Radisav

2008-01-01

Previous studies showed that temperature and total organic carbon in drinking water would cause chlorine dioxide (ClO(2)) loss in a water distribution system and affect the efficiency of ClO(2) for Legionella control. However, among the various causes of ClO(2) loss in a drinking water distribution system, the loss of disinfectant due to the reaction with corrosion scales has not been studied in detail. In this study, the corrosion scales from a galvanized iron pipe and a copper pipe that have been in service for more than 10 years were characterized by energy dispersive spectroscopy (EDS) and X-ray diffraction (XRD). The impact of these corrosion scale materials on ClO(2) decay was investigated in de-ionized water at 25 and 45 degrees C in a batch reactor with floating glass cover. ClO(2) decay was also investigated in a specially designed reactor made from the iron and copper pipes to obtain more realistic reaction rate data. Goethite (alpha-FeOOH) and magnetite (Fe(3)O(4)) were identified as the main components of iron corrosion scale. Cuprite (Cu(2)O) was identified as the major component of copper corrosion scale. The reaction rate of ClO(2) with both iron and copper oxides followed a first-order kinetics. First-order decay rate constants for ClO(2) reactions with iron corrosion scales obtained from the used service pipe and in the iron pipe reactor itself ranged from 0.025 to 0.083 min(-1). The decay rate constant for ClO(2) with Cu(2)O powder and in the copper pipe reactor was much smaller and it ranged from 0.0052 to 0.0062 min(-1). Based on these results, it can be concluded that the corrosion scale will cause much more significant ClO(2) loss in corroded iron pipes of the distribution system than the total organic carbon that may be present in finished water.

Heat pipes for use in a magnetic field

DOEpatents

Werner, Richard W.; Hoffman, Myron A.

1983-01-01

A heat pipe configuration for use in a magnetic field environment of a fusion reactor. Heat pipes for operation in a magnetic field when liquid metal working fluids are used are optimized by flattening of the heat pipes having an unobstructed annulus which significantly reduces the adverse side region effect of the prior known cylindrically configured heat pipes. The flattened heat pipes operating in a magnetic field can remove 2--3 times the heat as a cylindrical heat pipe of the same cross sectional area.

Maximal design basis accident of fusion neutron source DEMO-TIN

NASA Astrophysics Data System (ADS)

Kolbasov, B. N.

2015-12-01

When analyzing the safety of nuclear (including fusion) facilities, the maximal design basis accident at which the largest release of activity is expected must certainly be considered. Such an accident is usually the failure of cooling systems of the most thermally stressed components of a reactor (for a fusion facility, it is the divertor or the first wall). The analysis of safety of the ITER reactor and fusion power facilities (including hybrid fission-fusion facilities) shows that the initial event of such a design basis accident is a large-scale break of a pipe in the cooling system of divertor or the first wall outside the vacuum vessel of the facility. The greatest concern is caused by the possibility of hydrogen formation and the inrush of air into the vacuum chamber (VC) with the formation of a detonating mixture and a subsequent detonation explosion. To prevent such an explosion, the emergency forced termination of the fusion reaction, the mounting of shutoff valves in the cooling systems of the divertor and the first wall or blanket for reducing to a minimum the amount of water and air rushing into the VC, the injection of nitrogen or inert gas into the VC for decreasing the hydrogen and oxygen concentration, and other measures are recommended. Owing to a continuous feed-out of the molten-salt fuel mixture from the DEMO-TIN blanket with the removal period of 10 days, the radioactivity release at the accident will mainly be determined by tritium (up to 360 PBq). The activity of fission products in the facility will be up to 50 PBq.

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.

Analytical and experimental studies of heat pipe radiation cooling of hypersonic propulsion systems

NASA Technical Reports Server (NTRS)

Martin, R. A.; Merrigan, M. A.; Elder, M. G.; Sena, J. T.; Keddy, E. S.; Silverstein, C. C.

1992-01-01

Analytical and experimental studies were completed to assess the feasibility of using high-temperature heat pipes to cool hypersonic engine components. This new approach involves using heat pipes to transport heat away from the combustor, nozzle, or inlet regions, and to reject it to the environment by thermal radiation from an external heat pipe nacelle. For propulsion systems using heat pipe radiation cooling (HPRC), it is possible to continue to use hydrocarbon fuels into the Mach 4 to Mach 6 speed range, thereby enhancing the economic attractiveness of commercial or military hypersonic flight. In the second-phase feasibility program recently completed, it is found that heat loads produced by considering both convection and radiation heat transfer from the combustion gas can be handled with HPRC design modifications. The application of thermal insulation to ramburner and nozzle walls was also found to reduce the heat load by about one-half and to reduce peak HPRC system temperatures to below 2700 F. In addition, the operation of HPRC at cruise conditions of around Mach 4.5 and at an altitude of 90,000 ft lowers the peak hot-section temperatures to around 2800 F. An HPRC heat pipe was successfully fabricated and tested at Mach 5 conditions of heat flux, heat load, and temperature.

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

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.

Cooling Effect Analysis of Suppressing Coal Spontaneous Ignition with Heat Pipe

NASA Astrophysics Data System (ADS)

Zhang, Yaping; Zhang, Shuanwei; Wang, Jianguo; Hao, Gaihong

2018-05-01

Suppression of spontaneous ignition of coal stockpiles was an important issue for safe utilization of coal. The large thermal energy from coal spontaneous ignition can be viewed as the latent energy source to further utilize for saving energy purpose. Heat pipe was the more promising way to diffuse effectively concentrated energy of the coal stockpile, so that retarding coal spontaneous combustion was therefore highly desirable. The cooling mechanism of the coal with heat pipe was pursued. Based on the research result, the thermal energy can be transported from the coal seam to the surface continuously with the use of heat pipe. Once installed the heat pipes will work automatically as long as the coal oxidation reaction was happened. The experiment was indicated that it can significantly spread the high temperature of the coal pile.

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...

Design of the Sandia-Israel 20-kW reflux heat-pipe solar receiver/reactor

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

Diver, R.B.; Ginn, W.C.

1987-09-01

This report describes the design and fabrication of a 20-kW sodium reflux heat-pipe solar receiver/reactor for CO/sub 2/ reforming of methane. This project is part of a bilateral agreement between the United States and Israel. Under the terms of the agreement the solar receiver/reactor has been designed and built by Sandia National Laboratories for testing in the 7-meter solar furnace facility at the Weizmann Institute of Science in Rehovot, Israel. 16 refs., 11 figs., 2 tabs.

Altitude-wind-tunnel investigation of tail-pipe burning with a Westinghouse X24C-4B axial-flow turbojet engine

NASA Technical Reports Server (NTRS)

Fleming, William A; Wallner, Lewis E

1948-01-01

Thrust augmentation of an axial-flow type turbojet engine by burning fuel in the tail pipe has been investigated in the NACA Cleveland altitude wind tunnel. The performance was determined over a range of simulated flight conditions and tail-pipe fuel flows. The engine tail pipe was modified for the investigation to reduce the gas velocity at the inlet of the tail-pipe combustion chamber and to provide an adequate seat for the flame; four such modifications were investigated. The highest net-thrust increase obtained in the investigation was 86 percent with a net thrust specific fuel consumption of 2.91 and a total fuel-air ratio of 0.0523. The highest combustion efficiencies obtained with the four configurations ranged from 0.71 to 0.96. With three of the tail-pipe burners, for which no external cooling was provided, the exhaust nozzle and the rear part of the burner section were bright red during operation at high tail-pipe fuel-air ratios. With the tail-pipe burner for which fuel and water cooling were provided, the outer shell of the tail-pipe burner showed no evidence of elevated temperatures at any operating condition.

Lorentz force effect on mixed convection micropolar flow in a vertical conduit

NASA Astrophysics Data System (ADS)

Abdel-wahed, Mohamed S.

2017-05-01

The present work provides a simulation of control and filtration process of hydromagnetic blood flow with Hall current under the effect of heat source or sink through a vertical conduit (pipe). This work meets other engineering applications, such as nuclear reactors cooled during emergency shutdown, geophysical transport in electrically conducting and heat exchangers at low velocity conditions. The problem is modeled by a system of partial differential equations taking the effect of viscous dissipation, and these equations are simplified and solved analytically as a series solution using the Differential Transformation Method (DTM). The velocities and temperature profiles of the flow are plotted and discussed. Moreover, the conduit wall shear stress and heat flux are deduced and explained.

Comparative assessment of out-of-core nuclear thermionic power systems

NASA Technical Reports Server (NTRS)

Estabrook, W. C.; Koenig, D. R.; Prickett, W. Z.

1975-01-01

The hardware selections available for fabrication of a nuclear electric propulsion stage for planetary exploration were explored. The investigation was centered around a heat-pipe-cooled, fast-spectrum nuclear reactor for an out-of-core power conversion system with sufficient detail for comparison with the in-core system studies completed previously. A survey of competing power conversion systems still indicated that the modular reliability of thermionic converters makes them the desirable choice to provide the 240-kWe end-of-life power for at least 20,000 full power hours. The electrical energy will be used to operate a number of mercury ion bombardment thrusters with a specific impulse in the range of about 4,000-5,000 seconds.

Low-Cost Radiator for Fission Power Thermal Control

NASA Technical Reports Server (NTRS)

Maxwell, Taylor; Tarau, Calin; Anderson, William; Hartenstine, John; Stern, Theodore; Walmsley, Nicholas; Briggs, Maxwell

2014-01-01

NASA Glenn Research Center (GRC) is developing fission power system technology for future Lunar surface power applications. The systems are envisioned in the 10 to 100kW(sub e) range and have an anticipated design life of 8 to 15 years with no maintenance. NASA GRC is currently setting up a 55 kW(sub e) non-nuclear system ground test in thermal-vacuum to validate technologies required to transfer reactor heat, convert the heat into electricity, reject waste heat, process the electrical output, and demonstrate overall system performance. Reducing the radiator mass, size, and cost is essential to the success of the program. To meet these goals, Advanced Cooling Technologies, Inc. (ACT) and Vanguard Space Technologies, Inc. (VST) are developing a single facesheet radiator with heat pipes directly bonded to the facesheet. The facesheet material is a graphite fiber reinforced composite (GFRC) and the heat pipes are titanium/water. By directly bonding a single facesheet to the heat pipes, several heavy and expensive components can be eliminated from the traditional radiator design such as, POC(TradeMark) foam saddles, aluminum honeycomb, and a second facesheet. A two-heat pipe radiator prototype, based on the single facesheet direct-bond concept, was fabricated and tested to verify the ability of the direct-bond joint to withstand coefficient of thermal expansion (CTE) induced stresses during thermal cycling. The thermal gradients along the bonds were measured before and after thermal cycle tests to determine if the performance degraded. Overall, the results indicated that the initial uniformity of the adhesive was poor along one of the heat pipes. However, both direct bond joints showed no measureable amount of degradation after being thermally cycled at both moderate and aggressive conditions.

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

Kupca, L.; Beno, P.

A very brief summary is provided of a primary circuit piping material properties analysis. The analysis was performed for the Bohunice V-1 reactor and the Kola-1 and -2 reactors. Assessment was performed on Bohunice V-1 archive materials and primary piping material cut from the Kola units after 100,000 hours of operation. Main research program tasks included analysis of mechanical properties, corrosion stability, and microstructural properties. Analysis results are not provided.

63. REACTOR CHAMBER (BASF) FROM NORTH SHOWING NEUTRON SHIELD TANK ...

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

63. REACTOR CHAMBER (BASF) FROM NORTH SHOWING NEUTRON SHIELD TANK AND REACTOR PIPING (LOCATION RRR) - Shippingport Atomic Power Station, On Ohio River, 25 miles Northwest of Pittsburgh, Shippingport, Beaver County, PA

Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling

NASA Astrophysics Data System (ADS)

Domalapally, Phani; Di Caro, Marco

2018-05-01

Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.

Oscillating-Coolant Heat Exchanger

NASA Technical Reports Server (NTRS)

Scotti, Stephen J.; Blosser, Max L.; Camarda, Charles J.

1992-01-01

Devices useful in situations in which heat pipes inadequate. Conceptual oscillating-coolant heat exchanger (OCHEX) transports heat from its hotter portions to cooler portions. Heat transported by oscillation of single-phase fluid, called primary coolant, in coolant passages. No time-averaged flow in tubes, so either heat removed from end reservoirs on every cycle or heat removed indirectly by cooling sides of channels with another coolant. Devices include leading-edge cooling devices in hypersonic aircraft and "frost-free" heat exchangers. Also used in any situation in which heat pipe used and in other situations in which heat pipes not usable.

Flowpath evaluation and reconnaissance by remote field Eddy current testing (FERRET)

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

Smoak, A.E.; Zollinger, W.T.

1993-12-31

This document describes the design and development of FERRET (Flowpath Evaluation and Reconnaisance by Remote-field Eddy current Testing). FERRET is a system for inspecting the steel pipes which carry cooling water to underground nuclear waste storage tanks. The FERRET system has been tested in a small scale cooling pipe mock-up, an improved full scale mock-up, and in flaw detection experiments. Early prototype designs of FERRET and the FERRET launcher (a device which inserts, moves, and retrieves probes from a piping system) as well as the field-ready design are discussed.

Heat pipes for use in a magnetic field

DOEpatents

Werner, R.W.; Hoffman, M.A.

1983-07-19

A heat pipe configuration for use in a magnetic field environment of a fusion reactor is disclosed. Heat pipes for operation in a magnetic field when liquid metal working fluids are used are optimized by flattening of the heat pipes having an unobstructed annulus which significantly reduces the adverse side region effect of the prior known cylindrically configured heat pipes. The flattened heat pipes operating in a magnetic field can remove 2--3 times the heat as a cylindrical heat pipe of the same cross sectional area. 4 figs.

Heat pipe cooling of power processing magnetics

NASA Technical Reports Server (NTRS)

Hansen, I. G.; Chester, M.

1979-01-01

The constant demand for increased power and reduced mass has raised the internal temperature of conventionally cooled power magnetics toward the upper limit of acceptability. The conflicting demands of electrical isolation, mechanical integrity, and thermal conductivity preclude significant further advancements using conventional approaches. However, the size and mass of multikilowatt power processing systems may be further reduced by the incorporation of heat pipe cooling directly into the power magnetics. Additionally, by maintaining lower more constant temperatures, the life and reliability of the magnetic devices will be improved. A heat pipe cooled transformer and input filter have been developed for the 2.4 kW beam supply of a 30-cm ion thruster system. This development yielded a mass reduction of 40% (1.76 kg) and lower mean winding temperature (20 C lower). While these improvements are significant, preliminary designs predict even greater benefits to be realized at higher power. This paper presents the design details along with the results of thermal vacuum operation and the component performance in a 3 kW breadboard power processor.

Piping support system for liquid-metal fast-breeder reactor

DOEpatents

Brussalis, Jr., William G.

1984-01-01

A pipe support consisting of a rigid link pivotally attached to a pipe and an anchor, adapted to generate stress or strain in the link and pipe due to pipe thermal movement, which stress or strain can oppose further pipe movement and generally provides pipe support. The pipe support can be used in multiple combinations with other pipe supports to form a support system. This support system is most useful in applications in which the pipe is normally operated at a constant elevated or depressed temperature such that desired stress or strain can be planned in advance of pipe and support installation. The support system is therefore especially useful in steam stations and in refrigeration equipment.

75 FR 36698 - Draft Regulatory Guide: Issuance, Availability

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-28

... information based on the likelihood of pipe breaks of different sizes. The rule would divide all coolant... to and including a ``transition break size,'' and breaks larger than the transition size up to the largest pipe in the reactor coolant system. Selection of the transition size was based upon pipe break...

46 CFR 92.20-50 - Heating and cooling.

Code of Federal Regulations, 2010 CFR

2010-10-01

...) under normal operating conditions without curtailing ventilation. (c) Radiators and other heating... the occupants. Pipes leading to radiators or heating apparatus must be insulated where those pipes...

Passive cooling system for a vehicle

DOEpatents

Hendricks, Terry Joseph; Thoensen, Thomas

2005-11-15

A passive cooling system for a vehicle (114) transfers heat from an overheated internal component, for example, an instrument panel (100), to an external portion (116) of the vehicle (114), for example, a side body panel (126). The passive cooling system includes one or more heat pipes (112) having an evaporator section (118) embedded in the overheated internal component and a condenser section (120) at the external portion (116) of the vehicle (114). The evaporator (118) and condenser (120) sections are in fluid communication. The passive cooling system may also include a thermally conductive film (140) for thermally connecting the evaporator sections (118) of the heat pipes (112) to each other and to the instrument panel (100).

Passive Cooling System for a Vehicle

DOEpatents

Hendricks, T. J.; Thoensen, T.

2005-11-15

A passive cooling system for a vehicle (114) transfers heat from an overheated internal component, for example, an instrument panel (100), to an external portion (116) of the vehicle (114), for example, a side body panel (126). The passive cooling system includes one or more heat pipes (112) having an evaporator section (118) embedded in the overheated internal component and a condenser section (120) at the external portion (116) of the vehicle (114). The evaporator (118) and condenser (120) sections are in fluid communication. The passive cooling system may also include a thermally conductive film (140) for thermally connecting the evaporator sections (118) of the heat pipes (112) to each other and to the instrument panel (100).

Reactor Materials Program - Baseline Material Property Handbook - Mechanical Properties of 1950's Vintage Stainless Steel Weldment Components, Task Number 89-23-A-1

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

Stoner, K.J.

1999-11-05

The Process Water System (primary coolant) piping of the nuclear production reactors constructed in the 1950''s at Savannah River Site is comprised primarily of Type 304 stainless steel with Type 308 stainless steel weld filler. A program to measure the mechanical properties of archival PWS piping and weld materials (having approximately six years of service at temperatures between 25 and 100 degrees C) has been completed. The results from the mechanical testing has been synthesized to provide a mechanical properties database for structural analyses of the SRS piping.

Mitigating Climate Change with Ocean Pipes: Influencing Land Temperature and Hydrology and Termination Overshoot Risk

NASA Astrophysics Data System (ADS)

Kwiatkowski, L.; Caldeira, K.; Ricke, K.

2014-12-01

With increasing risk of dangerous climate change geoengineering solutions to Earth's climate problems have attracted much attention. One proposed geoengineering approach considers the use of ocean pipes as a means to increase ocean carbon uptake and the storage of thermal energy in the deep ocean. We use a latest generation Earth System Model (ESM) to perform simulations of idealised extreme implementations of ocean pipes. In our simulations, downward transport of thermal energy by ocean pipes strongly cools the near surface atmosphere - by up to 11°C on a global mean. The ocean pipes cause net thermal energy to be transported from the terrestrial environment to the deep ocean while increasing the global net transport of water to land. By cooling the ocean surface more than the land, ocean pipes tend to promote a monsoonal-type circulation, resulting in increased water vapour transport to land. Throughout their implementation, ocean pipes prevent energy from escaping to space, increasing the amount of energy stored in Earth's climate system despite reductions in surface temperature. As a consequence, our results indicate that an abrupt termination of ocean pipes could cause dramatic increases in surface temperatures beyond that which would have been obtained had ocean pipes not been implemented.

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.

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.

Gravity Scaling of a Power Reactor Water Shield

NASA Technical Reports Server (NTRS)

Reid, Robert S.; Pearson, J. Boise

2008-01-01

Water based reactor shielding is being considered as an affordable option for use on initial lunar surface power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxiliary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2007). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n). These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

Hydrothermal alteration of kimberlite by convective flows of external water.

PubMed

Afanasyev, A A; Melnik, O; Porritt, L; Schumacher, J C; Sparks, R S J

Kimberlite volcanism involves the emplacement of olivine-rich volcaniclastic deposits into volcanic vents or pipes. Kimberlite deposits are typically pervasively serpentinised as a result of the reaction of olivine and water within a temperature range of 130-400 °C or less. We present a model for the influx of ground water into hot kimberlite deposits coupled with progressive cooling and serpentisation. Large-pressure gradients cause influx and heating of water within the pipe with horizontal convergent flow in the host rock and along pipe margins, and upward flow within the pipe centre. Complete serpentisation is predicted for wide ranges of permeability of the host rocks and kimberlite deposits. For typical pipe dimensions, cooling times are centuries to a few millennia. Excess volume of serpentine results in filling of pore spaces, eventually inhibiting fluid flow. Fresh olivine is preserved in lithofacies with initial low porosity, and at the base of the pipe where deeper-level host rocks have low permeability, and the pipe is narrower leading to faster cooling. These predictions are consistent with fresh olivine and serpentine distribution in the Diavik A418 kimberlite pipe, (NWT, Canada) and with features of kimberlites of the Yakutian province in Russia affected by influx of ground water brines. Fast reactions and increases in the volume of solid products compared to the reactants result in self-sealing and low water-rock ratios (estimated at <0.2). Such low water-rock ratios result in only small changes in stable isotope compositions; for example, δO 18 is predicted only to change slightly from mantle values. The model supports alteration of kimberlites predominantly by interactions with external non-magmatic fluids.

46 CFR 32.40-50 - Heating and cooling-T/ALL.

Code of Federal Regulations, 2010 CFR

2010-10-01

...) under normal operating conditions without curtailing ventilation. (c) Radiators and other heating... the occupants. Pipes leading to radiators or heating apparatus must be insulated where those pipes...

Maximal design basis accident of fusion neutron source DEMO-TIN

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

Kolbasov, B. N., E-mail: Kolbasov-BN@nrcki.ru

2015-12-15

When analyzing the safety of nuclear (including fusion) facilities, the maximal design basis accident at which the largest release of activity is expected must certainly be considered. Such an accident is usually the failure of cooling systems of the most thermally stressed components of a reactor (for a fusion facility, it is the divertor or the first wall). The analysis of safety of the ITER reactor and fusion power facilities (including hybrid fission–fusion facilities) shows that the initial event of such a design basis accident is a large-scale break of a pipe in the cooling system of divertor or themore » first wall outside the vacuum vessel of the facility. The greatest concern is caused by the possibility of hydrogen formation and the inrush of air into the vacuum chamber (VC) with the formation of a detonating mixture and a subsequent detonation explosion. To prevent such an explosion, the emergency forced termination of the fusion reaction, the mounting of shutoff valves in the cooling systems of the divertor and the first wall or blanket for reducing to a minimum the amount of water and air rushing into the VC, the injection of nitrogen or inert gas into the VC for decreasing the hydrogen and oxygen concentration, and other measures are recommended. Owing to a continuous feed-out of the molten-salt fuel mixture from the DEMO-TIN blanket with the removal period of 10 days, the radioactivity release at the accident will mainly be determined by tritium (up to 360 PBq). The activity of fission products in the facility will be up to 50 PBq.« less

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

Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as thesemore » installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.« less

Progress of cryogenic pulsating heat pipes at UW-Madison

NASA Astrophysics Data System (ADS)

Diego Fonseca, Luis; Mok, Mason; Pfotenhauer, John; Miller, Franklin

2017-12-01

Space agencies continuously require innovative cooling systems that are lightweight, low powered, physically flexible, easily manufactured and, most importantly, exhibit high heat transfer rates. Therefore, Pulsating Heat Pipes (PHPs) are being investigated to provide these requirements. This paper summarizes the current development of cryogenic Pulsating Heat Pipes with single and multiple evaporator sections built and successfully tested at UW-Madison. Recently, a helium based Pulsating Heat Pipe with three evaporator and three condenser sections has been operated at fill ratios between 20 % and 80 % operating temperature range of 2.9 K to 5.19 K, resulting in a maximum effective thermal conductivity up to 50,000 W/m-K. In addition, a nitrogen Pulsating Heat Pipe has been built with three evaporator sections and one condenser section. This PHP achieved a thermal performance between 32,000 W/m-K and 96,000 W/m-K at fill ratio ranging from 50 % to 80 %. Split evaporator sections are very important in order to spread cooling throughout an object of interest with an irregular temperature distribution or where multiple cooling locations are required. Hence this type of configurations is a proof of concept which hasn’t been attempted before and if matured could be applied to cryo-propellant tanks, superconducting magnets and photon detectors.

Heat Pipe Planets

NASA Astrophysics Data System (ADS)

Moore, W. B.; Simon, J. I.

2018-05-01

We propose that cooling via volcanic heat pipes may provide a universal model of the way terrestrial bodies transition from a magma-ocean state into subsequent single-plate, stagnant-lid convection or plate tectonic phases.

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.

Magnetic refrigeration apparatus with heat pipes

DOEpatents

Barclay, John A.; Prenger, Jr., F. Coyne

1987-01-01

A magnetic refrigerator operating in the 4 to 20 K range utilizes heat pipes to transfer heat to and from the magnetic material at the appropriate points during the material's movement. In one embodiment circular disks of magnetic material can be interleaved with the ends of the heat pipes. In another embodiment a mass of magnetic material reciprocatingly moves between the end of the heat pipe of pipes that transmits heat from the object of cooling to the magnetic material and the end of the heat pipe or pipes that transmits heat from the magnetic material to a heat sink.

Magnetic refrigeration apparatus with heat pipes

DOEpatents

Barclay, J.A.; Prenger, F.C. Jr.

1985-10-25

A magnetic refrigerator operating in the 4 to 20 K range utilizes heat pipes to transfer heat to and from the magnetic material at the appropriate points during the material's movement. In one embodiment circular disks of magnetic material can be interleaved with the ends of the heat pipes. In another embodiment a mass of magnetic material reciprocatingly moves between the end of the heat pipe or pipes that transmits heat from the object of cooling to the magnetic material and the end of the heat pipe or pipes that transmits heat from the magnetic material to a heat sink.

Demonstration and Validation of Corrosion-Mitigation Technologies for Mechanical Room Utility Piping and Cooling-Tower Pumps

DTIC Science & Technology

2015-05-01

Infrastructure, Task 2.1” ERDC/CERL TR-15-5 ii Abstract Two critical infrastructure corrosion issues at Fort Bragg, NC, are the cor- rosion of steel utility...piping union joints in mechanical rooms and the cor- rosion of steel pump housings in cooling tower systems. Reliable operation of these components...pump 5 incorporating 316 stainless steel housing. .................................... 19 Figure 13. New pump 5 being installed

Effect of Structure Factor on High-Temperature Ductility of Pipe Steels

NASA Astrophysics Data System (ADS)

Kolbasnikov, N. G.; Matveev, M. A.; Mishnev, P. A.

2016-05-01

Effects of various factors such as the grain size, the morphology of nonmetallic inclusions, and joint microalloying with boron and titanium on the high-temperature ductility of pipe steels are studied. Physical modeling of the conditions of cooling of the skin of a continuous-cast preform in the zone of secondary cooling in a Gleeble facility is performed. Technical recommendations are given for raising the hot ductility of steels under industrial conditions.

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)

Remaining Sites Verification Package for the 100-F-26:12, 1.8-m (72-in.) Main Process Sewer Pipeline, Waste Site Reclassification Form 2007-034

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

J. M. Capron

2008-04-29

The 100-F-26:12 waste site was an approximately 308-m-long, 1.8-m-diameter east-west-trending reinforced concrete pipe that joined the North Process Sewer Pipelines (100-F-26:1) and the South Process Pipelines (100-F-26:4) with the 1.8-m reactor cooling water effluent pipeline (100-F-19). In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

Optimize out-of-core thermionic energy conversion for nuclear electric propulsion

NASA Technical Reports Server (NTRS)

Morris, J. F.

1978-01-01

Thermionic energy conversion (TEC) potentialities for nuclear electric propulsion (NEP) are examined. Considering current designs, their limitations, and risks raises critical questions about the use of TEC for NEP. Apparently a reactor cooled by hotter-than-1675 K heat pipes has good potentialities. TEC with higher temperatures and greater power densities than the currently proposed 1650 K, 5-to-6 W/sq cm version offers substantial gains. Other approaches to high-temperature electric isolation appear also promising. A high-power-density, high-temperature TEC for NEP appears, therefore, attainable. It is recommended to optimize out-of-core thermionic energy conversion for nuclear electric propulsion. Although current TEC designs for NEP seem unnecessary compared with Brayton versions, large gains are apparently possible with increased temperatures and greater power densities.

Transient thermohydraulic heat pipe modeling

NASA Astrophysics Data System (ADS)

Hall, Michael L.; Doster, Joseph M.

Many space based reactor designs employ heat pipes as a means of conveying heat. In these designs, thermal radiation is the principle means for rejecting waste heat from the reactor system, making it desirable to operate at high temperatures. Lithium is generally the working fluid of choice as it undergoes a liquid-vapor transformation at the preferred operating temperature. The nature of remote startup, restart, and reaction to threats necessitates an accurate, detailed transient model of the heat pipe operation. A model is outlined of the vapor core region of the heat pipe which is part of a large model of the entire heat pipe thermal response. The vapor core is modeled using the area averaged Navier-Stokes equations in one dimension, which take into account the effects of mass, energy and momentum transfer. The core model is single phase (gaseous), but contains two components: lithium gas and a noncondensible vapor. The vapor core model consists of the continuity equations for the mixture and noncondensible, as well as mixture equations for internal energy and momentum.

Heat pipe with embedded wick structure

DOEpatents

Adkins, Douglas Ray; Shen, David S.; Tuck, Melanie R.; Palmer, David W.; Grafe, V. Gerald

1998-01-01

A heat pipe has an embedded wick structure that maximizes capillary pumping capability. Heat from attached devices such as integrated circuits evaporates working fluid in the heat pipe. The vapor cools and condenses on a heat dissipation surface. The condensate collects in the wick structure, where capillary pumping returns the fluid to high heat areas.

Heat pipe with embedded wick structure

DOEpatents

Adkins, D.R.; Shen, D.S.; Tuck, M.R.; Palmer, D.W.; Grafe, V.G.

1998-06-23

A heat pipe has an embedded wick structure that maximizes capillary pumping capability. Heat from attached devices such as integrated circuits evaporates working fluid in the heat pipe. The vapor cools and condenses on a heat dissipation surface. The condensate collects in the wick structure, where capillary pumping returns the fluid to high heat areas. 7 figs.

Heat pipe with embedded wick structure

DOEpatents

Adkins, Douglas Ray; Shen, David S.; Tuck, Melanie R.; Palmer, David W.; Grafe, V. Gerald

1999-01-01

A heat pipe has an embedded wick structure that maximizes capillary pumping capability. Heat from attached devices such as integrated circuits evaporates working fluid in the heat pipe. The vapor cools and condenses on a heat dissipation surface. The condensate collects in the wick structure, where capillary pumping returns the fluid to high heat areas.

Development and study of a heat pipe with dielectric properties

NASA Astrophysics Data System (ADS)

Semena, M. G.; Gershuni, A. N.; Chepurnoi, A. B.

Requirements for the structural elements of heat pipes with dielectric properties are examined. To obtain information necessary for the thermal analysis of heat pipes, a study is made of the capillary-transport characteristics of a dielectric capillary structure consisting of quartz fibers; the capillary pressure and the liquid penetration coefficient are determined. The results of the study are used to develop dielectric heat pipes for the cooling of a vacuum electronic instrument. Experimentally determined characteristics of the heat pipes are presented.

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.

Rigorous derivation of the effective model describing a non-isothermal fluid flow in a vertical pipe filled with porous medium

NASA Astrophysics Data System (ADS)

Beneš, Michal; Pažanin, Igor

2018-03-01

This paper reports an analytical investigation of non-isothermal fluid flow in a thin (or long) vertical pipe filled with porous medium via asymptotic analysis. We assume that the fluid inside the pipe is cooled (or heated) by the surrounding medium and that the flow is governed by the prescribed pressure drop between pipe's ends. Starting from the dimensionless Darcy-Brinkman-Boussinesq system, we formally derive a macroscopic model describing the effective flow at small Brinkman-Darcy number. The asymptotic approximation is given by the explicit formulae for the velocity, pressure and temperature clearly acknowledging the effects of the cooling (heating) and porous structure. The theoretical error analysis is carried out to indicate the order of accuracy and to provide a rigorous justification of the effective model.

46 CFR 182.425 - Engine exhaust cooling.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 46 Shipping 7 2012-10-01 2012-10-01 false Engine exhaust cooling. 182.425 Section 182.425 Shipping...) MACHINERY INSTALLATION Specific Machinery Requirements § 182.425 Engine exhaust cooling. (a) Except as otherwise provided in this paragraph, all engine exhaust pipes must be water cooled. (1) Vertical dry...

46 CFR 182.425 - Engine exhaust cooling.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 46 Shipping 7 2013-10-01 2013-10-01 false Engine exhaust cooling. 182.425 Section 182.425 Shipping...) MACHINERY INSTALLATION Specific Machinery Requirements § 182.425 Engine exhaust cooling. (a) Except as otherwise provided in this paragraph, all engine exhaust pipes must be water cooled. (1) Vertical dry...

46 CFR 182.425 - Engine exhaust cooling.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 46 Shipping 7 2011-10-01 2011-10-01 false Engine exhaust cooling. 182.425 Section 182.425 Shipping...) MACHINERY INSTALLATION Specific Machinery Requirements § 182.425 Engine exhaust cooling. (a) Except as otherwise provided in this paragraph, all engine exhaust pipes must be water cooled. (1) Vertical dry...

46 CFR 182.425 - Engine exhaust cooling.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 46 Shipping 7 2014-10-01 2014-10-01 false Engine exhaust cooling. 182.425 Section 182.425 Shipping...) MACHINERY INSTALLATION Specific Machinery Requirements § 182.425 Engine exhaust cooling. (a) Except as otherwise provided in this paragraph, all engine exhaust pipes must be water cooled. (1) Vertical dry...

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

Aluminum/ammonia heat pipe gas generation and long term system impact for the Space Telescope's Wide Field Planetary Camera

NASA Technical Reports Server (NTRS)

Jones, J. A.

1983-01-01

In the Space Telescope's Wide Field Planetary Camera (WFPC) project, eight heat pipes (HPs) are used to remove heat from the camera's inner electronic sensors to the spacecraft's outer, cold radiator surface. For proper device functioning and maximization of the signal-to-noise ratios, the Charge Coupled Devices (CCD's) must be maintained at -95 C or lower. Thermoelectric coolers (TEC's) cool the CCD's, and heat pipes deliver each TEC's nominal six to eight watts of heat to the space radiator, which reaches an equilibrium temperature between -15 C to -70 C. An initial problem was related to the difficulty to produce gas-free aluminum/ammonia heat pipes. An investigation was, therefore, conducted to determine the cause of the gas generation and the impact of this gas on CCD cooling. In order to study the effect of gas slugs in the WFPC system, a separate HP was made. Attention is given to fabrication, testing, and heat pipe gas generation chemistry studies.

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.

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Technical Reports Server (NTRS)

Martin, James; Salvail, Pat

2003-01-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final "wet in". A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/- 1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to less than10(exp -10) std cc/sec helium and vacuum conditioned at 250 C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Astrophysics Data System (ADS)

Martin, James; Salvail, Pat

2004-02-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final ``wet in''. A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/-1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to <10-10 std cc/sec helium and vacuum conditioned at 250 °C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 °C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

Nuclear reactor heat transport system component low friction support system

DOEpatents

Wade, Elman E.

1980-01-01

A support column for a heavy component of a liquid metal fast breeder reactor heat transport system which will deflect when the pipes leading coolant to and from the heavy component expand or contract due to temperature changes includes a vertically disposed pipe, the pipe being connected to the heavy component by two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles and the pipe being supported through two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles, each of the cylindrical surfaces bearing on a flat and horizontal surface.

PARTIAL ECONOMIC STUDY OF STEAM COOLED HEAVY WATER MODERATED REACTORS

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

None

1960-04-01

Steam-cooled reactors are compared with CAHDU for costs of Calandria tubes, pressure tubes. heavy water moderator, heavy water reflector, fuel supply, heat exchanger, and turbine generator. A direct-cycle lightsteam-cooled heavy- water-moderated pressure-tube reactor formed the basic reactor design for the study. Two methods of steam circulation through the reactor were examined. In both cases the steam was generated outside the reactor and superheated in the reactor core. One method consisted of a series of reactor and steam generator passes. The second method consisted of the Loeffler cycle and its modifications. The fuel was assumed to be natural cylindrical UO/sub 2/more » pellets sheathed in a hypothetical material with the nuclear properties of Zircaloy, but able to function at temperatures to 900 deg F. For the conditions assumed, the longer the rod, the higher the outlet temperature and therefore the higher the efficiency. The turbine cycle efficiency was calculated on the assumption that suitable steam generators are available. As the neutron losses to the pressure tubes were significant, an economic analysis of insulated pressure tubes is included. A description of the physics program for steam-cooled reactors is included. Results indicated that power from the steam-cooled reactor would cost 1.4 mills/ kwh compared with 1.25 mills/kwh for CANDU. (M.C.G.)« less

System Design for a Nuclear Electric Spacecraft Utilizing Out-of-core Thermionic Conversion

NASA Technical Reports Server (NTRS)

Estabrook, W. C.; Phillips, W. M.; Hsieh, T.

1976-01-01

Basic guidelines are presented for a nuclear space power system which utilizes heat pipes to transport thermal power from a fast nuclear reactor to an out of core thermionic converter array. Design parameters are discussed for the nuclear reactor, heat pipes, thermionic converters, shields (neutron and gamma), waste heat rejection systems, and the electrical bus bar-cable system required to transport the high current/low voltage power to the processing equipment. Dimensions are compatible with shuttle payload bay constraints.

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

Goetsch, D.; Bieniussa, K.; Schulz, H.

This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branchingmore » pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.« less

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.

Characteristics of a 1.6 W Gifford-McMahon Cryocooler with a Double Pipe Regenerator

NASA Astrophysics Data System (ADS)

Masuyama, S.; Numazawa, T.

2017-12-01

This paper focuses on the second stage regenerator of a 4 K Gifford-McMahon (G-M) cryocooler. A three-layer layout of lead (Pb), HoCu2 and Gd2O2S spheres in the second stage regenerator derives a good performance at 4 K. After some modifications, we confirmed that the cooling power of 1.60 W at 4.2 K was achieved by using this three-layer layout. A two-stage G-M cryocooler is RDK-408D2 (SHI) and a compressor is C300G (SUZUKISHOKAN) with a rated electric input power of 7.3 kW at 60 Hz. In order to further improve, a double pipe regenerator was applied to the second stage regenerator. As a double pipe, a stainless steel pipe with thin wall was inserted in the coaxial direction into the second stage regenerator. The helium flow in the second stage regenerator is expected to be non-uniform flow because of the distribution of helium density and the imperfect packing of regenerator material. The double pipe regenerator is considered to have an effect of restraining the non-uniform flow. From the experimental results, the second stage cooling power of 1.67 W at 4.2 K and the first stage cooling power of 64.9 W at 50 K were achieved.

150. ARAIII Reactor building (ARA608) Sections. Show highbay section, heater ...

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

150. ARA-III Reactor building (ARA-608) Sections. Show high-bay section, heater stack, and depth of reactor, piping, and heater pits. Aerojet-general 880-area/GCRE-608-A-3. Date: February 1958. Ineel index code no. 063-0608-00-013-102613. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Heat pipe cooling of power processing magnetics

NASA Technical Reports Server (NTRS)

Hansen, I. G.; Chester, M. S.

1979-01-01

A heat pipe cooled transformer and input filter were developed for the 2.4 kW beam supply of a 30 cm ion thruster system. This development yielded a mass reduction of 40% (1.76 kg) and lower mean winding temperature (20 C lower). While these improvements are significant, preliminary designs predict even greater benefits to be realized at higher power. The design details are presented along with the results of thermal vacuum operation and the component performance in a 3 kW breadboard power processor.

ERTS-C (Landsat 3) cryogenic heat pipe experiment definition

NASA Technical Reports Server (NTRS)

Brennan, P. J.; Kroliczek, E. J.

1975-01-01

A flight experiment designed to demonstrate current cryogenic heat pipe technology was defined and evaluated. The experiment package developed is specifically configured for flight aboard an ERTS type spacecraft. Two types of heat pipes were included as part of the experiment package: a transporter heat pipe and a thermal diode heat pipe. Each was tested in various operating modes. Performance data obtained from the experiment are applicable to the design of cryogenic systems for detector cooling, including applications where periodic high cooler temperatures are experienced as a result of cyclic energy inputs.

Heat Pipe Planets

NASA Technical Reports Server (NTRS)

Moore, William B.; Simon, Justin I.; Webb, A. Alexander G.

2014-01-01

When volcanism dominates heat transport, a terrestrial body enters a heat-pipe mode, in which hot magma moves through the lithosphere in narrow channels. Even at high heat flow, a heat-pipe planet develops a thick, cold, downwards-advecting lithosphere dominated by (ultra-)mafic flows and contractional deformation at the surface. Heat-pipes are an important feature of terrestrial planets at high heat flow, as illustrated by Io. Evidence for their operation early in Earth's history suggests that all terrestrial bodies should experience an episode of heat-pipe cooling early in their histories.

Engineering design aspects of the heat-pipe power system

NASA Technical Reports Server (NTRS)

Capell, B. M.; Houts, M. G.; Poston, D. I.; Berte, M.

1997-01-01

The Heat-pipe Power System (HPS) is a near-term, low-cost space power system designed at Los Alamos that can provide up to 1,000 kWt for many space nuclear applications. The design of the reactor is simple, modular, and adaptable. The basic design allows for the use of a variety of power conversion systems and reactor materials (including the fuel, clad, and heat pipes). This paper describes a project that was undertaken to develop a database supporting many engineering aspects of the HPS design. The specific tasks discussed in this paper are: the development of an HPS materials database, the creation of finite element models that will allow a wide variety of investigations, and the verification of past calculations.

A theoretical and computational study of lithium-ion battery thermal management for electric vehicles using heat pipes

NASA Astrophysics Data System (ADS)

Greco, Angelo; Cao, Dongpu; Jiang, Xi; Yang, Hong

2014-07-01

A simplified one-dimensional transient computational model of a prismatic lithium-ion battery cell is developed using thermal circuit approach in conjunction with the thermal model of the heat pipe. The proposed model is compared to an analytical solution based on variable separation as well as three-dimensional (3D) computational fluid dynamics (CFD) simulations. The three approaches, i.e. the 1D computational model, analytical solution, and 3D CFD simulations, yielded nearly identical results for the thermal behaviours. Therefore the 1D model is considered to be sufficient to predict the temperature distribution of lithium-ion battery thermal management using heat pipes. Moreover, a maximum temperature of 27.6 °C was predicted for the design of the heat pipe setup in a distributed configuration, while a maximum temperature of 51.5 °C was predicted when forced convection was applied to the same configuration. The higher surface contact of the heat pipes allows a better cooling management compared to forced convection cooling. Accordingly, heat pipes can be used to achieve effective thermal management of a battery pack with confined surface areas.

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

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)

Experiment data report for Semiscale Mod-1 Test S-05-1 (alternate ECC injection test)

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

Feldman, E. M.; Patton, Jr., M. L.; Sackett, K. E.

Recorded test data are presented for Test S-05-1 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-1 was conducted from initial conditions of 2263 psia and 544/sup 0/F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the vessel lower plenum to simulatemore » emergency core coolant injection in a PWR, with the flow rate based on system volume scaling.« less

Recent Advances in Power Conversion and Heat Rejection Technology for Fission Surface Power

NASA Technical Reports Server (NTRS)

Mason, Lee

2010-01-01

Under the Exploration Technology Development Program, the National Aeronautics and Space Administration (NASA) and the Department of Energy (DOE) are jointly developing Fission Surface Power (FSP) technology for possible use in human missions to the Moon and Mars. A preliminary reference concept was generated to guide FSP technology development. The concept consists of a liquid-metal-cooled reactor, Stirling power conversion, and water heat rejection, with Brayton power conversion as a backup option. The FSP project has begun risk reduction activities on some key components with the eventual goal of conducting an end-to-end, non-nuclear, integrated system test. Several power conversion and heat rejection hardware prototypes have been built and tested. These include multi-kilowatt Stirling and Brayton power conversion units, titanium-water heat pipes, and composite radiator panels.

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.

Heat pipes in space and on earth

NASA Technical Reports Server (NTRS)

Ollendorf, S.

1978-01-01

The performance of heat pipes used in the thermal control system of spacecraft such as OAO-III and ATS-6 is discussed, and applications of heat pipes to permafrost stabilization on the Alaska Pipeline and to heat recovery systems are described. Particular attention is given to the ATS-6, launched in 1974, which employs 55 heat pipes to carry solar and internal power loads to radiator surfaces. In addition, experiments involving radiative cooling based on cryogenic heat pipes have been planned for the Long Duration Exposure Facility spacecraft and for Spacelab. The role of heat pipes in Space Shuttle heat rejection services is also mentioned.

Detail exterior view looking north showing piping system adjacent to ...

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

Detail exterior view looking north showing piping system adjacent to engine house. Gas cooling system is on far right. - Burnsville Natural Gas Pumping Station, Saratoga Avenue between Little Kanawha River & C&O Railroad line, Burnsville, Braxton County, WV

Modeling of transient heat pipe operation

NASA Technical Reports Server (NTRS)

Colwell, Gene T.

1989-01-01

Mathematical models and an associated computer program for heat pipe startup from the frozen state have been developed. Finite element formulations of the governing equations are written for each heat pipe region for each operating condition during startup from the frozen state. The various models were checked against analytical and experimental data available in the literature for three specific types of operation. Computations using the methods developed were made for a space shuttle reentry mission where a heat pipe cooled leading edge was used on the wing.

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

7. COOLING TOWER FROM ROOF. Hot Springs National Park, ...

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

7. COOLING TOWER FROM ROOF. - Hot Springs National Park, Bathhouse Row, Quapaw Bathhouse: Mechanical & Piping Systems, State Highway 7, 1 mile north of U.S. Highway 70, Hot Springs, Garland County, AR

Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

DOEpatents

Schreiber, R.B.; Fero, A.H.; Sejvar, J.

1997-12-16

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 to form 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 valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

DOEpatents

Schreiber, Roger B.; Fero, Arnold H.; Sejvar, James

1997-01-01

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 to form 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 valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

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.

PBF Cooling Tower Auxiliary Building (PER624) interior. Camera facing north. ...

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

PBF Cooling Tower Auxiliary Building (PER-624) interior. Camera facing north. Deluge valves and automatic fire protection piping for Cooling Tower. Photographer: Holmes. Date: May 20, 1970. INEEL negative no. 70-2323 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Visualisation of flow patterns in straight and C-shape thermosyphons

NASA Astrophysics Data System (ADS)

Ong, K. S.; Tshai, K. H.; Firwana, A.

2017-04-01

A heat pipe is a passive heat transfer device capable of transferring a large quantity of heat effectively and efficiently over a long distance and with a small temperature difference between the heat source and heat sink. A heat pipe consists of a metal pipe initially vacuumed and then filled with a small quantity of fluid inside. The pipe is separated into a heating (evaporator) section and a cooling (condenser) section by an adiabatic section. In a run-around-coil heating, ventilation and air conditioning system, a wrap-around heat pipe heat exchanger could be employed to increase dehumidification and to reduce cooling costs. The thermal performance of a thermosyphon is dependent upon type of fill liquid, fill ratio, power input, pipe inclination and pipe dimensions. The boiling and condensation processes that occur inside a thermosyphon are quite complex. During operation, dry-out, burn-out or boiling limit, entrainment or flooding limit and geysering occur. These phenomena would lead to non-uniform axial wall temperature distribution in the pipe, or worse still, ineffective operation. In order to have a better understanding of the internal heat transfer phenomena, a visual study using transparent glass tubes and high speed camera recording of the internal flow patterns would be most helpful. This paper reports on an experimental investigation conducted to visualise the flow patterns in straight and C-shape thermosyphons. The pictures recorded enabled the internal flow boiling and condensation pattern occurring inside a straight and a C-shape thermosyphon to be observed. The thermosyphons were fabricated from 10 mm O/D × 8 mm I/D × 300 mm long glass tubes and filled with water with fill ratios from 0.5 - 1.5. The evaporator sections of the thermosyphons were immersed into a hot water tank that was electrically heated from cold at ambient temperature till boiling. Cooling of the condenser section was achieved using a fan. Preliminary results showed that dry-out occurred earlier at lower evaporator temperatures with small fill ratios. Further investigations to determine saturation and thermosyphon wall temperatures with various fill liquids and at different fill ratios, inclinations and pipe sizes are necessary with a more sophisticated video recording system.

Seismic Design of ITER Component Cooling Water System-1 Piping

NASA Astrophysics Data System (ADS)

Singh, Aditya P.; Jadhav, Mahesh; Sharma, Lalit K.; Gupta, Dinesh K.; Patel, Nirav; Ranjan, Rakesh; Gohil, Guman; Patel, Hiren; Dangi, Jinendra; Kumar, Mohit; Kumar, A. G. A.

2017-04-01

The successful performance of ITER machine very much depends upon the effective removal of heat from the in-vessel components and other auxiliary systems during Tokamak operation. This objective will be accomplished by the design of an effective Cooling Water System (CWS). The optimized piping layout design is an important element in CWS design and is one of the major design challenges owing to the factors of large thermal expansion and seismic accelerations; considering safety, accessibility and maintainability aspects. An important sub-system of ITER CWS, Component Cooling Water System-1 (CCWS-1) has very large diameter of pipes up to DN1600 with many intersections to fulfill the process flow requirements of clients for heat removal. Pipe intersection is the weakest link in the layout due to high stress intensification factor. CCWS-1 piping up to secondary confinement isolation valves as well as in-between these isolation valves need to survive a Seismic Level-2 (SL-2) earthquake during the Tokamak operation period to ensure structural stability of the system in the Safe Shutdown Earthquake (SSE) event. This paper presents the design, qualification and optimization of layout of ITER CCWS-1 loop to withstand SSE event combined with sustained and thermal loads as per the load combinations defined by ITER and allowable limits as per ASME B31.3, This paper also highlights the Modal and Response Spectrum Analyses done to find out the natural frequency and system behavior during the seismic event.

Conduction-driven cooling of LED-based automotive LED lighting systems for abating local hot spots

NASA Astrophysics Data System (ADS)

Saati, Ferina; Arik, Mehmet

2018-02-01

Light-emitting diode (LED)-based automotive lighting systems pose unique challenges, such as dual-side packaging (front side for LEDs and back side for driver electronics circuit), size, harsh ambient, and cooling. Packaging for automotive lighting applications combining the advanced printed circuit board (PCB) technology with a multifunctional LED-based board is investigated with a focus on the effect of thermal conduction-based cooling for hot spot abatement. A baseline study with a flame retardant 4 technology, commonly known as FR4 PCB, is first compared with a metal-core PCB technology, both experimentally and computationally. The double-sided advanced PCB that houses both electronics and LEDs is then investigated computationally and experimentally compared with the baseline FR4 PCB. Computational models are first developed with a commercial computational fluid dynamics software and are followed by an advanced PCB technology based on embedded heat pipes, which is computationally and experimentally studied. Then, attention is turned to studying different heat pipe orientations and heat pipe placements on the board. Results show that conventional FR4-based light engines experience local hot spots (ΔT>50°C) while advanced PCB technology based on heat pipes and thermal spreaders eliminates these local hot spots (ΔT<10°C), leading to a higher lumen extraction with improved reliability. Finally, possible design options are presented with embedded heat pipe structures that further improve the PCB performance.

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.

Vaporization Would Cool Primary Battery

NASA Technical Reports Server (NTRS)

Bhandari, Pradeep; Miyake, Robert N.

1991-01-01

Temperature of discharging high-power-density primary battery maintained below specified level by evaporation of suitable liquid from jacket surrounding battery, according to proposal. Pressure-relief valve regulates pressure and boiling temperature of liquid. Less material needed in cooling by vaporization than in cooling by melting. Technique used to cool batteries in situations in which engineering constraints on volume, mass, and location prevent attachment of cooling fins, heat pipes, or like.

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.

90. ARAIII. GCRE reactor building (ARA608) mechanical loop pit. Shows ...

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

90. ARA-III. GCRE reactor building (ARA-608) mechanical loop pit. Shows nitrogen gas compressor in foreground, piping installations on walls of pit, and other details. February 24, 1959. Ineel photo no. 59-880. Photographer: Ken Mansfield. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

10. NEEDLE SHOWER IN COOLING ROOM. Hot Springs National ...

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

10. NEEDLE SHOWER IN COOLING ROOM. - Hot Springs National Park, Bathhouse Row, Fordyce Bathhouse: Mechanical & Piping Systems, State Highway 7, 1 mile north of U.S. Highway 70, Hot Springs, Garland County, AR

6. UNIT VENTILATOR, WOMEN'S COOLING ROOM. Hot Springs National ...

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

6. UNIT VENTILATOR, WOMEN'S COOLING ROOM. - Hot Springs National Park, Bathhouse Row, Ozark Bathhouse: Mechanical & Piping Systems, State Highway 7, 1 mile north of U.S. Highway 70, Hot Springs, Garland County, AR

Solar heating and cooling systems design and development

NASA Technical Reports Server (NTRS)

1976-01-01

Solar heating and heating/cooling systems were designed for single family, multifamily, and commercial applications. Subsystems considered included solar collectors, heat storage systems, auxiliary energy sources, working fluids, and supplementary controls, piping, and pumps.

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.

Depriming of arterial heat pipes: An investigation of CTS thermal excursions

NASA Technical Reports Server (NTRS)

Antoniuk, D.; Edwards, D. K.

1980-01-01

Four thermal excursions of the Transmitter Experiment Package (TEP) were the result of the depriming of the arteries in all three heat pipes in the Variable Conductance Heat Pipe System which cooled the TEP. The determined cause of the depriming of the heat pipes was the formation of bubbles of the nitrogen/helium control gas mixture in the arteries during the thaw portion of a freeze/thaw cycle of the inactive region of the condenser section of the heat pipe. Conditions such as suction freezeout or heat pipe turn-on, which moved these bubbles into the active region of the heat pipe, contributed to the depriming mechanism. Methods for precluding, or reducing the probability of, this type of failure mechanism in future applications of arterial heat pipes are included.

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.

Fabrication and Testing of a Leading-Edge-Shaped Heat Pipe

NASA Technical Reports Server (NTRS)

Glass, David E.; Merrigan, Michael A.; Sena, J. Tom; Reid, Robert S.

1998-01-01

The development of a refractory-composite/heat-pipe-cooled leading edge has evolved from the design stage to the fabrication and testing of a full size, leading-edge-shaped heat pipe. The heat pipe had a 'D-shaped' cross section and was fabricated from arc cast Mo-4lRe. An artery was included in the wick. Several issues were resolved with the fabrication of the sharp leading edge radius heat pipe. The heat pipe was tested in a vacuum chamber at Los Alamos National Laboratory using induction heating and was started up from the frozen state several times. However, design temperatures and heat fluxes were not obtained due to premature failure of the heat pipe resulting from electrical discharge between the induction heating apparatus and the heat pipe. Though a testing anomaly caused premature failure of the heat pipe, successful startup and operation of the heat pipe was demonstrated.

Analysis of BF Hearth Reasonable Cooling System Based on the Water Dynamic Characteristics

NASA Astrophysics Data System (ADS)

Zuo, Haibin; Jiao, Kexin; Zhang, Jianliang; Li, Qian; Wang, Cui

A rational cooling water system is the assurance for long campaign life of blast furnace. In the paper, the heat transfer of different furnace period and different furnace condition based on the water quality characteristics were analysed, and the reason of the heat flux over the normal from the hydrodynamics was analysed. The results showed that, the vapour-film and scale existence significantly influenced the hearth heat transfer, which accelerated the brick lining erosion. The water dynamic characteristics of the parallel inner pipe or among the pipes were the main reason for the abnormal heat flux and film boiling. As to the reasonable cooling water flow, the gas film and the scale should be controlled and the energy saving should be considered.

Passive ice freezing-releasing heat pipe

DOEpatents

Gorski, Anthony J.; Schertz, William W.

1982-01-01

A heat pipe device has been developed which permits completely passive ice formation and periodic release of ice without requiring the ambient temperature to rise above the melting point of water. This passive design enables the maximum amount of cooling capacity to be stored in the tank.

Design characteristics of a heat pipe test chamber

NASA Technical Reports Server (NTRS)

Baker, Karl W.; Jang, J. Hoon; Yu, Juin S.

1992-01-01

LeRC has designed a heat pipe test facility which will be used to provide data for validating heat pipe computer codes. A heat pipe test chamber that uses helium gas for enhancing heat transfer was investigated. The conceptual design employs the technique of guarded heating and guarded cooling to facilitate accurate measurements of heat transfer rates to the evaporator and from the condenser. The design parameters are selected for a baseline heat pipe made of stainless steel with an inner diameter of 38.10 mm and a wall thickness of 1.016 mm. The heat pipe operates at a design temperature of 1000 K with an evaporator radial heat flux of 53 W/sq. cm.

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

Nuclear reactor system study for NASA/JPL

NASA Technical Reports Server (NTRS)

Palmer, R. G.; Lundberg, L. B.; Keddy, E. S.; Koenig, D. R.

1982-01-01

Reactor shielding, safety studies, and heat pipe development work are described. Monte Carlo calculations of gamma and neutron shield configurations show that substantial weight penalties are incurred if exposure at 25 m to neutrons and gammas must be limited to 10 to the 12th power nvt and 10 to the 6th power rad, instead of the 10 to the 13th power nvt and 10 to the 7th power rad values used earlier. For a 1.6 MW sub t reactor, the required shield weight increases from 400 to 815 kg. Water immersion critically calculations were extended to study the effect of water in fuel void spaces as well as in the core heat pipes. These show that the insertion into the core of eight blades of B4C with a mass totaling 2.5 kg will guarantee subcriticality. The design, fabrication procedure, and testing of a 4m long molybdenum/lithium heat pipe are described. It appears that an excess of oxygen in the wick prevented the attainment of expected performance capability.

46 CFR 56.85-5 - Heating and cooling method.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 46 Shipping 2 2011-10-01 2011-10-01 false Heating and cooling method. 56.85-5 Section 56.85-5 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING PIPING SYSTEMS AND APPURTENANCES Heat Treatment of Welds § 56.85-5 Heating and cooling method. Heat treatment may be accomplished...

46 CFR 56.85-5 - Heating and cooling method.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Heating and cooling method. 56.85-5 Section 56.85-5 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING PIPING SYSTEMS AND APPURTENANCES Heat Treatment of Welds § 56.85-5 Heating and cooling method. Heat treatment may be accomplished...

46 CFR 56.85-5 - Heating and cooling method.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 46 Shipping 2 2014-10-01 2014-10-01 false Heating and cooling method. 56.85-5 Section 56.85-5 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING PIPING SYSTEMS AND APPURTENANCES Heat Treatment of Welds § 56.85-5 Heating and cooling method. Heat treatment may be accomplished...

46 CFR 56.85-5 - Heating and cooling method.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 46 Shipping 2 2013-10-01 2013-10-01 false Heating and cooling method. 56.85-5 Section 56.85-5 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING PIPING SYSTEMS AND APPURTENANCES Heat Treatment of Welds § 56.85-5 Heating and cooling method. Heat treatment may be accomplished...

46 CFR 56.85-5 - Heating and cooling method.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 46 Shipping 2 2012-10-01 2012-10-01 false Heating and cooling method. 56.85-5 Section 56.85-5 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING PIPING SYSTEMS AND APPURTENANCES Heat Treatment of Welds § 56.85-5 Heating and cooling method. Heat treatment may be accomplished...

Initial evaluation of floor cooling on lactating sows under severe acute heat stress

USDA-ARS?s Scientific Manuscript database

The objective was to evaluate the effects of floor cooling on lactating sows under severe summer heat stress. Twelve multiparous sows were provided with a cooling pad built with an aluminum plate surface, high-density polyethylene base and copper pipes. Treatments were randomly allotted to sows to r...

Effect of floor cooling on late lactation sows under acute heat stress

USDA-ARS?s Scientific Manuscript database

The objective was to evaluate the effects of floor cooling on late lactation sows under severe summer heat stress. Ten multiparous sows were provided with a cooling pad built with an aluminum plate surface, high-density polyethylene base and copper pipes. Treatments were randomly allotted to sows to...

Effects of floor cooling on late lactation sows under severe acute heat stress

USDA-ARS?s Scientific Manuscript database

The objective was to evaluate the effects of floor cooling on late lactation sows under severe summer heat stress. Ten multiparous sows were provided with a cooling pad built with an aluminum plate surface, high-density polyethylene base and copper pipes. Treatments were randomly allotted to sows to...

Initial evaluation of floor cooling on lactating sows under severe acute heat stress

USDA-ARS?s Scientific Manuscript database

The objectives were to evaluate an acute heat stress protocol for lactating sows and evaluate preliminary estimates of water flow rates required to cool sows. Twelve multiparous sows were provided with a cooling pad built with an aluminum plate surface, high-density polyethylene base and copper pipe...

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 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.

PBF Cooling Tower. Hot deck of Cooling Tower with fan ...

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

PBF Cooling Tower. Hot deck of Cooling Tower with fan motors in place. Fan's propeller blades (not in view) rotate within lower portion of vents. Inlet pipe is a left of view. Contractor's construction buildings in view to right. Photographer: Larry Page. Date: June 30, 1969. INEEL negative no. 69-3781 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Passive ice freezing-releasing heat pipe. [Patent application

DOEpatents

Gorski, A.J.; Schertz, W.W.

1980-09-29

A heat pipe device has been developed which permits completely passive ice formation and periodic release of ice without requiring the ambient temperature to rise above the melting point of water. This passive design enables the maximum amount of cooling capacity to be stored in the tank.

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

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.

Preliminary Design of a SP-100/Stirling Radiatively Coupled Heat Exchanger

NASA Technical Reports Server (NTRS)

Schmitz, Paul; Tower, Leonard; Dawson, Ronald; Blue, Brian; Dunn, Pat

1995-01-01

Several methods for coupling the SP-100 space nuclear reactor to the NASA Lewis Research Center's Free Piston Stirling Power Convertor (FPSPC) are presented. A 25 kWe, dual opposed Stirling convertor configuration is used in these designs. The concepts use radiative coupling between the SP-100 lithium loop and the sodium heat pipe of the Stirling convertor to transfer the heat from the reactor to the convertor. Four separate configurations are presented. Masses for the four designs vary from 41 to 176 kgs. Each design's structure, heat transfer characteristics, and heat pipe performance are analytically modeled.

Preliminary design of a SP-100/Stirling radiatively coupled heat exchanger

NASA Astrophysics Data System (ADS)

Schmitz, Paul; Tower, Leonard; Dawson, Ronald; Blue, Brian; Dunn, Pat

1995-10-01

Several methods for coupling the SP-100 space nuclear reactor to the NASA Lewis Research Center's Free Piston Stirling Power Convertor (FPSPC) are presented. A 25 kWe, dual opposed Stirling convertor configuration is used in these designs. The concepts use radiative coupling between the SP-100 lithium loop and the sodium heat pipe of the Stirling convertor to transfer the heat from the reactor to the convertor. Four separate configurations are presented. Masses for the four designs vary from 41 to 176 kgs. Each design's structure, heat transfer characteristics, and heat pipe performance are analytically modeled.

Efficiently-cooled plasmonic amorphous silicon solar cells integrated with a nano-coated heat-pipe plate

PubMed Central

Zhang, Yinan; Du, Yanping; Shum, Clifford; Cai, Boyuan; Le, Nam Cao Hoai; Chen, Xi; Duck, Benjamin; Fell, Christopher; Zhu, Yonggang; Gu, Min

2016-01-01

Solar photovoltaics (PV) are emerging as a major alternative energy source. The cost of PV electricity depends on the efficiency of conversion of light to electricity. Despite of steady growth in the efficiency for several decades, little has been achieved to reduce the impact of real-world operating temperatures on this efficiency. Here we demonstrate a highly efficient cooling solution to the recently emerging high performance plasmonic solar cell technology by integrating an advanced nano-coated heat-pipe plate. This thermal cooling technology, efficient for both summer and winter time, demonstrates the heat transportation capability up to ten times higher than those of the metal plate and the conventional wickless heat-pipe plates. The reduction in temperature rise of the plasmonic solar cells operating under one sun condition can be as high as 46%, leading to an approximate 56% recovery in efficiency, which dramatically increases the energy yield of the plasmonic solar cells. This newly-developed, thermally-managed plasmonic solar cell device significantly extends the application scope of PV for highly efficient solar energy conversion. PMID:27113558

Efficiently-cooled plasmonic amorphous silicon solar cells integrated with a nano-coated heat-pipe plate.

PubMed

Zhang, Yinan; Du, Yanping; Shum, Clifford; Cai, Boyuan; Le, Nam Cao Hoai; Chen, Xi; Duck, Benjamin; Fell, Christopher; Zhu, Yonggang; Gu, Min

2016-04-26

Solar photovoltaics (PV) are emerging as a major alternative energy source. The cost of PV electricity depends on the efficiency of conversion of light to electricity. Despite of steady growth in the efficiency for several decades, little has been achieved to reduce the impact of real-world operating temperatures on this efficiency. Here we demonstrate a highly efficient cooling solution to the recently emerging high performance plasmonic solar cell technology by integrating an advanced nano-coated heat-pipe plate. This thermal cooling technology, efficient for both summer and winter time, demonstrates the heat transportation capability up to ten times higher than those of the metal plate and the conventional wickless heat-pipe plates. The reduction in temperature rise of the plasmonic solar cells operating under one sun condition can be as high as 46%, leading to an approximate 56% recovery in efficiency, which dramatically increases the energy yield of the plasmonic solar cells. This newly-developed, thermally-managed plasmonic solar cell device significantly extends the application scope of PV for highly efficient solar energy conversion.

Efficiently-cooled plasmonic amorphous silicon solar cells integrated with a nano-coated heat-pipe plate

NASA Astrophysics Data System (ADS)

Zhang, Yinan; Du, Yanping; Shum, Clifford; Cai, Boyuan; Le, Nam Cao Hoai; Chen, Xi; Duck, Benjamin; Fell, Christopher; Zhu, Yonggang; Gu, Min

2016-04-01

Solar photovoltaics (PV) are emerging as a major alternative energy source. The cost of PV electricity depends on the efficiency of conversion of light to electricity. Despite of steady growth in the efficiency for several decades, little has been achieved to reduce the impact of real-world operating temperatures on this efficiency. Here we demonstrate a highly efficient cooling solution to the recently emerging high performance plasmonic solar cell technology by integrating an advanced nano-coated heat-pipe plate. This thermal cooling technology, efficient for both summer and winter time, demonstrates the heat transportation capability up to ten times higher than those of the metal plate and the conventional wickless heat-pipe plates. The reduction in temperature rise of the plasmonic solar cells operating under one sun condition can be as high as 46%, leading to an approximate 56% recovery in efficiency, which dramatically increases the energy yield of the plasmonic solar cells. This newly-developed, thermally-managed plasmonic solar cell device significantly extends the application scope of PV for highly efficient solar energy conversion.

Heat Pipes and Heat Rejection Component Testing at NASA Glenn Research Center

NASA Technical Reports Server (NTRS)

Sanzi, James L.; Jaworske, Donald A.

2012-01-01

Titanium-water heat pipes are being evaluated for use in the heat rejection system for space fission power systems. The heat rejection syst em currently comprises heat pipes with a graphite saddle and a composite fin. The heat input is a pumped water loop from the cooling of the power conversion system. The National Aeronautics and Space Administration has been life testing titanium-water heat pipes as well as eval uating several heat pipe radiator designs. The testing includes thermal modeling and verification of model, material compatibility, frozen startup of heat pipe radiators, and simulating low-gravity environments. Future thermal testing of titanium-water heat pipes includes low-g ravity testing of thermosyphons, radiation testing of heat pipes and fin materials, water pump performance testing, as well as Small Busine ss Innovation Research funded deliverable prototype radiator panels.

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.

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.

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.

A Small Fission Power System for NASA Planetary Science Missions

NASA Technical Reports Server (NTRS)

Mason, Lee; Casani, John; Elliott, John; Fleurial, Jean-Pierre; MacPherson, Duncan; Nesmith, William; Houts, Michael; Bechtel, Ryan; Werner, James; Kapernick, Rick;

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.

Cooling Acoustic Transducer with Heat Pipes

DTIC Science & Technology

2009-07-29

a heat sink. [0009] In Kan et al (United States Patent No. 6,528,909), a spindle motor assembly is disclosed which has a shaft with an integral...heat pipe. The shaft with the integral heat pipe improves the thermal conductively of the shaft and the spindle motor assembly. The shaft includes...Description of the Prior Art [0004] It is known in the art that transducers, designed to project acoustic power, are often limited by the build

PRELIMINARY HAZARDS SUMMARY REPORT FOR THE VALLECITOS SUPERHEAT REACTOR

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

Murray, J.L.

1961-02-01

BS>The Vallecitos Superheat Reactor (VSR) is a light-watermoderated, thermal-spectrum reactor, cooled by a combination of moderator boiling and forced convection cooling with saturated steam. The reactor core consists of 32 fuel hurdles containing 5300 lb of UO/sub 2/ enriched in U/sub 235/ to 3.6%. The fuel elements are arranged in individual process tubes that direct the cooling steam flow and separate the steam from the water moderator. The reactor vessel is designed for 1250 psig and operates at 960 to 1000 psig. With the reactor operating at 12.5 Mw(t), the maximum fuel cladding temperature is 1250 deg F and themore » cooling steam is superheated to an average temperature of about 810 deg F at 905 psig. Nu clear operation of the reactor is controlled by 12 control rods, actuated by drives mounted on the bottom of the reactor vessel. The water moderator recirculates inside the reactor vessel and through the core region by natural convection. Inherent safety features of the reactor include the negative core reactivity effects upon heating the UO/sub 2/ fuel (Doppler effect), upon increasing the temperature or void content of the moderator in the operating condition, and upon unflooding the fuel process tubes in the hot condition. Snfety features designed into the reactor and plant systems include a system of sensors and devices to detect petentially unsafe operating conditions and to initiate automatically the appropriate countermeasures, a set of fast and reliable control rods for scramming the reactor if a potentially unsafe condition occurs, a manually-actuated liquid neutron poison system, and an emergency cooling system to provide continued steam flow through the reactor core in the event the reactor becomes isolated from either its normal source of steam supply or discharge. The release of radioactivity to unrestricted areas is maintained within permissible limits by monitoring the radioactivity of wastes and controlling their release. The reactor and many of its auxiliaries are housed within a high-integrity essentially leak-tight containment vessel. (auth)« less

Non-Nuclear Testing of Compact Reactor Technologies at NASA MSFC

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Pearson, J. Boise; Godfroy, Thomas J.

2011-01-01

Safe, reliable, compact, autonomous, long-life fission systems have numerous potential applications, both terrestrially and in space. Technologies and facilities developed in support of these systems could be useful to a variety of concepts. At moderate power levels, fission systems can be designed to operate for decades without the need for refueling. In addition, fast neutron damage to cladding and structural materials can be maintained at an acceptable level. Nuclear design codes have advanced to the stage where high confidence in the behavior and performance of a system can be achieved prior to initial testing. To help ensure reactor affordability, an optimal strategy must be devised for development and qualification. That strategy typically involves a combination of non-nuclear and nuclear testing. Non-nuclear testing is particularly useful for concepts in which nuclear operating characteristics are well understood and nuclear effects such as burnup and radiation damage are not likely to be significant. To be mass efficient, a SFPS must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial reactors. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference SFPS uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance. In addition, technologies from the SFPS system could be applicable to compact terrestrial systems. Recent non-nuclear testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference SFPS and evaluate methods for system integration. In July, 2011 an Annular Linear Induction Pump (ALIP) provided by Idaho National Laboratory was tested at the EFF-TF to assess performance and verify suitability for use in a10 kWe technology demonstration unit (TDU). In November, 2011 testing of a 37-pin core simulator (designed in conjunction with Los Alamos National Laboratory) for use with the TDU will occur. Previous testing at the EFFTF has included the thermal and mechanical coupling of a pumped NaK loop to Stirling engines (provided by GRC). Testing related to heat pipe cooled systems, gas cooled systems, heat exchangers, and other technologies has also been performed. Integrated TDU testing will begin at GRC in 2013. Thermal simulators developed at the EFF-TF are capable of operating over the temperature and power range typically of interest to compact reactors. Small and large diameter simulators have been developed, and simulators (coupled with the facility) are able to closely match the axial and radial power profile of all potential systems of interest. A photograph of the TDU core simulator during assembly is provided in Figure 2.

Design of a Hydrogen Pulsating Heat Pipe

NASA Astrophysics Data System (ADS)

Liu, Yumeng; Deng, Haoren; Pfotenhauer, John; Gan, Zhihua

In order to enhance the application of a cryocooler that provides cooling capacity at the cold head location, and effectively spread that cooling over an extended region, one requires an efficient heat transfer method. The pulsating heat pipe affords a highly effective heat transfer component that has been extensively researched at room temperature, but is recently being investigated for cryogenic applications. This paper describes the design. The experimental setup is designed to characterize the thermal performance of the PHP as a function of the applied heat, number of turns, filling ratio, inclination angle, and length of adiabatic section.

Nuclear Design of the HOMER-15 Mars Surface Fission Reactor

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

Poston, David I.

2002-07-01

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

ETR HEAT EXCHANGER BUILDING, TRA644. WORKERS ARE INSTALLING HEAT EXCHANGER ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. WORKERS ARE INSTALLING HEAT EXCHANGER PIPING. INL NEGATIVE NO. 56-3122. Jack L. Anderson, Photographer, 9/21/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Long-term ice storage for cooling applications

DOEpatents

Schertz, William W.

1981-01-01

A device is providing for cooling a stored material and then for later use of the cold thus stored. The device includes a tank containing a liquid such as water which is frozen by means of a reflux condenser heat pipe.

Long-term ice storage for cooling applications

DOEpatents

Schertz, W.W.

A device is described for cooling a stored material and then for later use of the cold thus stored. The device includes a tank containing a liquid such as water which is frozen by means of a reflux condenser heat pipe.

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.

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.

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.

Spaceborne power systems preference analyses. Volume 2: Decision analysis

NASA Technical Reports Server (NTRS)

Smith, J. H.; Feinberg, A.; Miles, R. F., Jr.

1985-01-01

Sixteen alternative spaceborne nuclear power system concepts were ranked using multiattribute decision analysis. The purpose of the ranking was to identify promising concepts for further technology development and the issues associated with such development. Four groups were interviewed to obtain preference. The four groups were: safety, systems definition and design, technology assessment, and mission analysis. The highest ranked systems were the heat-pipe thermoelectric systems, heat-pipe Stirling, in-core thermionic, and liquid-metal thermoelectric systems. The next group contained the liquid-metal Stirling, heat-pipe Alkali Metal Thermoelectric Converter (AMTEC), heat-pipe Brayton, liquid-metal out-of-core thermionic, and heat-pipe Rankine systems. The least preferred systems were the liquid-metal AMTEC, heat-pipe thermophotovoltaic, liquid-metal Brayton and Rankine, and gas-cooled Brayton. The three nonheat-pipe technologies selected matched the top three nonheat-pipe systems ranked by this study.

Cooling Acoustic Transcucer with Heat Pipes

DTIC Science & Technology

2009-07-19

circuits to a heat sink. [0009] In Kan et al (United States Patent No. 6,528,909), a spindle motor assembly is disclosed which has a shaft with an...integral heat pipe. The shaft with the integral heat pipe improves the thermal conductively of the shaft and the spindle motor assembly. The shaft ...2) Description of the Prior Art [0004] It is known in the art that transducers, designed to project acoustic power, are often limited by the

Biofilm thickness measurement using an ultrasound method in a liquid phase.

PubMed

Maurício, R; Dias, C J; Jubilado, N; Santana, F

2013-10-01

In this report, the development of an online, noninvasive, measurement method of the biofilm thickness in a liquid phase is presented. The method is based in the analysis of the ultrasound wave pulse-echo behavior in a liquid phase reproducing the real reactor conditions. It does not imply the removal of the biomass from the support or any kind of intervention in the support (pipes) to detect and perform the measurements (non-invasiveness). The developed method allows for its sensor to be easily and quickly mounted and unmounted in any location along a pipe or reactor wall. Finally, this method is an important innovation because it allows the thickness measurement of a biofilm, in liquid phase conditions that can be used in monitoring programs, to help in scheduling cleaning actions to remove the unwanted biofilm, in several application areas, namely in potable water supply pipes.

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

LaSalle, F.R.; Golbeg, P.R.; Chenault, D.M.

For reactor and nuclear facilities, both Title 10, Code of Federal Regulations, Part 50, and US Department of Energy Order 6430.1A require assessments of the interaction of non-Safety Class 1 piping and equipment with Safety Class 1 piping and equipment during a seismic event to maintain the safety function. The safety class systems of nuclear reactors or nuclear facilities are designed to the applicable American Society of Mechanical Engineers standards and Seismic Category 1 criteria that require rigorous analysis, construction, and quality assurance. Because non-safety class systems are generally designed to lesser standards and seismic criteria, they may become missilesmore » during a safe shutdown earthquake. The resistance of piping, tubing, and equipment to seismically generated missiles is addressed in the paper. Gross plastic and local penetration failures are considered with applicable test verification. Missile types and seismic zones of influence are discussed. Field qualification data are also developed for missile evaluation.« less

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

Vins, M.

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe.more » Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)« less

Long titanium heat pipes for high-temperature space radiators

NASA Technical Reports Server (NTRS)

Girrens, S. P.; Ernst, D. M.

1982-01-01

Titanium heat pipes are being developed to provide light weight, reliable heat rejection devices as an alternate radiator design for the Space Reactor Power System (SP-100). The radiator design includes 360 heat pipes, each of which is 5.2 m long and dissipates 3 kW of power at 775 K. The radiator heat pipes use potassium as the working fluid, have two screen arteries for fluid return, a roughened surface distributive wicking system, and a D shaped cross section container configuration. A prototype titanium heat pipe, 5.5 m long, was fabricated and tested in space simulating conditions. Results from startup and isothermal operation tests are presented. These results are also compared to theoretical performance predictions that were used to design the heat pipe initially.

Long titanium heat pipes for high-temperature space radiators

NASA Technical Reports Server (NTRS)

Girrens, S. P.; Ernst, D. M.

1982-01-01

Titanium heat pipes are being developed to provide light weight, reliable heat rejection devices as an alternate radiator design for the Space Reactor Power System (SP-100). The radiator design includes 360 heat pipes, each of which is 5.2 m long and dissipates 3 kW of power at 775 K. The radiator heat pipes use potassium as the working fluid, have two screen arteries for fluid return, a roughened surface distributive wicking system, and a D-shaped cross-section container configuration. A prototype titanium heat pipe, 5.5-m long, has been fabricated and tested in space-simulating conditions. Results from startup and isothermal operation tests are presented. These results are also compared to theoretical performance predictions that were used to design the heat pipe initially.

Modeling of a CeO2 thermochemistry reduction process for hydrogen production by solar concentrated energy

NASA Astrophysics Data System (ADS)

Valle-Hernández, Julio; Romero-Paredes, Hernando; Arancibia-Bulnes, Camilo A.; Villafan-Vidales, Heidi I.; Espinosa-Paredes, Gilberto

2016-05-01

In this paper the simulation of the thermal reduction for hydrogen production through the decomposition of cerium oxide is presented. The thermochemical cycle for hydrogen production consists of the endothermic reduction of CeO2 at high temperature, where concentrated solar energy is used as a source of heat; and of the subsequent steam hydrolysis of the resulting cerium oxide to produce hydrogen. For the thermochemical process, a solar reactor prototype is proposed; consisting of a cubic receptacle made of graphite fiber thermally insulated. Inside the reactor a pyramidal arrangement with nine tungsten pipes is housed. The pyramidal arrangement is made respect to the focal point where the reflected energy is concentrated. The solar energy is concentrated through the solar furnace of high radiative flux. The endothermic step is the reduction of the cerium oxide to lower-valence cerium oxide, at very high temperature. The exothermic step is the hydrolysis of the cerium oxide (III) to form H2 and the corresponding initial cerium oxide made at lower temperature inside the solar reactor. For the modeling, three sections of the pipe where the reaction occurs were considered; the carrier gas inlet, the porous medium and the reaction products outlet. The mathematical model describes the fluid mechanics; mass and energy transfer occurring therein inside the tungsten pipe. Thermochemical process model was simulated in CFD. The results show a temperature distribution in the solar reaction pipe and allow obtaining the fluid dynamics and the heat transfer within the pipe. This work is part of the project "Solar Fuels and Industrial Processes" from the Mexican Center for Innovation in Solar Energy (CEMIE-Sol).

Lamp for generating high power ultraviolet radiation

DOEpatents

Morgan, Gary L.; Potter, James M.

2001-01-01

The apparatus is a gas filled ultraviolet generating lamp for use as a liquid purifier. The lamp is powred by high voltage AC, but has no metallic electrodes within or in contact with the gas enclosure which is constructed as two concentric quartz cylinders sealed together at their ends with the gas fill between the cylinders. Cooling liquid is pumped through the volume inside the inner quartz cylinder where an electrically conductive pipe spaced from the inner cylinder is used to supply the cooling liquid and act as the high voltage electrode. The gas enclosure is enclosed within but spaced from a metal housing which is connected to operate as the ground electrode of the circuit and through which the treated fluid flows. Thus, the electrical circuit is from the central pipe, and through the cooling liquid, the gas enclosure, the treated liquid on the outside of the outer quartz cylinder, and to the housing. The high voltage electrode is electrically isolated from the source of cooling liquid by a length of insulated hose which also supplies the cooling liquid.

Microstructure characterization and charpy toughness of P91 weldment for as-welded, post-weld heat treatment and normalizing & tempering heat treatment

NASA Astrophysics Data System (ADS)

Pandey, Chandan; Mahapatra, M. M.; Kumar, Pradeep; Giri, A.

2017-09-01

The effect of weld groove design and heat treatment on microstructure evolution and Charpy toughness of P91 pipe weldments was studied. The P91 pipe weldments were subjected to subcritical post weld heat treatment (760 °C-2 h) and normalizing/tempering conditions (normalized-1040 °C/40 min, air cooled; tempered 760 °C/2 h, air cooled) were employed. The influence of subsequent PWHT and N&T treatment on the microstructure of various zone of P91 pipe weldments were also investigated. The present investigation also described the effect of PWHT and N&T treatment on hardness, grain size, precipitate size, inter-particle spacing and fraction area of precipitates present in each zone of P91 pipe weldments. The result indicated great impact of heat treatment on the Charpy toughness and microstructure evolution of P91 weldments. The N&T treatment was found to be more effective heat treatment compared to subsequent PWHT. Charpy toughness value was found to be higher for narrow-groove design as compared to conventional V-groove design.

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.

MTR WING, TRA604, INTERIOR. BASEMENT. INTERIOR VIEW FROM SAME LOCATION ...

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

MTR WING, TRA-604, INTERIOR. BASEMENT. INTERIOR VIEW FROM SAME LOCATION IN WEST CORRIDOR AS PHOTO ID-33-G-42 BUT CAMERA FACES SOUTH. SIGN ON DOOR FOR "PIPE TUNNEL" WARNS OF RADIOLOGICAL AND ASBESTOS HAZARDS. DOOR HAS METAL HASPS. SIGN ON OVERHEAD WASTE HEAT RECOVERY PIPES SAYS THEY CONTAIN "ASBESTOS FREE INSULATION." FIRE DOOR AT LEFT LEADS TO STAIRWAY TO FIRST FLOOR. DOOR AT RIGHT LEADS TO ROOM WHICH ONCE CONTAINED MTR LIBRARY. INL NEGATIVE NO. HD46-13-4. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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.

Special Purpose Nuclear Reactor (5 MW) for Reliable Power at Remote Sites Assessment Report

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

Sterbentz, James William; Werner, James Elmer; McKellar, Michael George

The Phenomena Identification and Ranking Table (PIRT) technique was conducted on the Special Purpose Reactor nuclear plant design. The PIRT is a structured process to identify safety-relevant/safety-significant phenomena and assess the importance and knowledge base by ranking the phenomena. The Special Purpose Reactor is currently in the conceptual design stage. The candidate reactor has a solid monolithic stainless steel core with an array of heat pipes and fuel pellets embedded in the monolith. The heat pipes are used to remove heat from the core using simple, reliable, and well-characterized physics (capillarity, boiling, and condensation). In the initial design, one heatmore » exchanger is used for the working fluid that produces energy, and a second heat exchanger is used to remove decay heat in emergency or shutdown conditions. In addition, a power conversion cycle such as an open-air Brayton system is available as an option for power conversion and process heat. This report summarizes and documents the process and scope of the four PIRT reviews, noting the major activities and conclusions. The identified phenomena, analyses, rationales, and associated ratings are presented along with a summary of the findings from the four individual PIRTs, namely (1) Reactor Accident and Normal Operations, (2) Heat Pipes, (3) Materials, and (4) Power Conversion. The PIRT reports for these four major system areas evaluated are attached as appendixes to this report and provide considerably more detail about each assessment as well as a more complete listing of the phenomena that were evaluated.« less

ETR HEAT EXCHANGER BUILDING, TRA644. FLOOR PLAN AND SECTIONS. PUMP ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. FLOOR PLAN AND SECTIONS. PUMP CUBICLES WITH PUMP MOTORS OUTSIDE CUBICLES. HEAT EXCHANGER EQUIPMENT. COOLANT PIPE TUNNEL ENTERS FROM REACTOR BUILDING. KAISER ETR-5582-MTR-644-A-3, 2/1956. INL INDEX NO. 532-0644-00-486-101294, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642. NORTHSOUTH SECTION, LOOKING WEST. STEELFRAME ROOF, CRANE RAIL, ...

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

ETR, TRA-642. NORTH-SOUTH SECTION, LOOKING WEST. STEEL-FRAME ROOF, CRANE RAIL, AND CRANES. COOLANT PIPE TUNNEL LEADING TO REACTOR FROM EAST. (THIS WAS A PRELIMINARY CONCEPT DRAWING.) KAISER ETR-5528-MTR-642-A-4, 11/1955. INL INDEX NO. 532-0642-00-486-100912, REV. 1. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Cool pool development. Quarterly technical report No. 1, April-June 1979

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

Crowther, K.

1979-10-15

The Cool Pool is a passive cooling system consisting of a shaded, evaporating roof pond which thermosiphons cool water into water-filled, metal columns (culvert pipes) located within the building living space. The water in the roof pond is cooled by evaporation, convection and radiation. Because the water in the pool and downcomer is colder and denser than the water in the column a pressure difference is created and the cold water flows from the pool, through the downcomer and into the bottom of the column. The warm column water rises and flows through a connecting pipe into the pool. Itmore » is then cooled and the cycle repeats itself. The system requires no pumps. The water column absorbs heat from the building interior primarily by convection and radiation. Since the column is radiating at a significantly lower temperature than the interior walls it plays a double role in human comfort. Not only does it cool the air by convection but it provides a heat sink to which people can radiate. Since thermal radiation is important to the cooling of people, the cold water column contributes substantially to their feelings of comfort. Research on the Cool Pool system includes the following major tasks: control of biological organisms and debris in the roof pond and water cylinders; development of a heat exchanger; experimental investigation of the system's thermal performance; and development of a predictive computer simulation of the Cool Pool. Progress in these tasks is reported.« 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.

Methods for calculating conjugate problems of heat transfer

NASA Astrophysics Data System (ADS)

Kalinin, E. K.; Dreitser, G. A.; Kostiuk, V. V.; Berlin, I. I.

Methods are examined for calculating various conjugate problems of heat transfer in channels and closed vessels in cases of single-phase and two-phase flow in steady and unsteady conditions. The single-phase-flow studies involve the investigation of gaseous and liquid heat-carriers in pipes, annular and plane channels, and pipe bundles in cases of cooling and heating. General relationships are presented for heat transfer in cases of film, transition, and nucleate boiling, as well as for boiling crises. Attention is given to methods for analyzing the filling and cooling of conduits and tanks by cryogenic liquids; and ways to intensify heat transfer in these conditions are examined.

Investigations on the heat transport capability of a cryogenic oscillating heat pipe and its application in achieving ultra-fast cooling rates for cell vitrification cryopreservation☆

PubMed Central

Han, Xu; Ma, Hongbin; Jiao, Anjun; Critser, John K.

2010-01-01

Theoretically, direct vitrification of cell suspensions with relatively low concentrations (~1 M) of permeating cryoprotective agents (CPA) is suitable for cryopreservation of almost all cell types and can be accomplished by ultra-fast cooling rates that are on the order of 106–7 K/min. However, the methods and devices currently available for cell cryopreservation cannot achieve such high cooling rates. In this study, we constructed a novel cryogenic oscillating heat pipe (COHP) using liquid nitrogen as its working fluid and investigated its heat transport capability to assess its application for achieving ultra-fast cooling rates for cell cryopreservation. The experimental results showed that the apparent heat transfer coefficient of the COHP can reach 2 × 105 W/m2·K, which is two orders of the magnitude higher than traditional heat pipes. Theoretical analyzes showed that the average local heat transfer coefficient in the thin film evaporation region of the COHP can reach 1.2 × 106 W/m2·K, which is approximately 103 times higher than that achievable with standard pool-boiling approaches. Based on these results, a novel device design applying the COHP and microfabrication techniques is proposed and its efficiency for cell vitrification is demonstrated through numerical simulation. The estimated average cooling rates achieved through this approach is 106–7 K/min, which is much faster than the currently available methods and sufficient for achieving vitrification with relatively low concentrations of CPA. PMID:18430413

Investigations on the heat transport capability of a cryogenic oscillating heat pipe and its application in achieving ultra-fast cooling rates for cell vitrification cryopreservation.

PubMed

Han, Xu; Ma, Hongbin; Jiao, Anjun; Critser, John K

2008-06-01

Theoretically, direct vitrification of cell suspensions with relatively low concentrations ( approximately 1 M) of permeating cryoprotective agents (CPA) is suitable for cryopreservation of almost all cell types and can be accomplished by ultra-fast cooling rates that are on the order of 10(6-7) K/min. However, the methods and devices currently available for cell cryopreservation cannot achieve such high cooling rates. In this study, we constructed a novel cryogenic oscillating heat pipe (COHP) using liquid nitrogen as its working fluid and investigated its heat transport capability to assess its application for achieving ultra-fast cooling rates for cell cryopreservation. The experimental results showed that the apparent heat transfer coefficient of the COHP can reach 2 x 10(5) W/m(2).K, which is two orders of the magnitude higher than traditional heat pipes. Theoretical analyzes showed that the average local heat transfer coefficient in the thin film evaporation region of the COHP can reach 1.2 x 10(6) W/m(2).K, which is approximately 10(3) times higher than that achievable with standard pool-boiling approaches. Based on these results, a novel device design applying the COHP and microfabrication techniques is proposed and its efficiency for cell vitrification is demonstrated through numerical simulation. The estimated average cooling rates achieved through this approach is 10(6-7)K/min, which is much faster than the currently available methods and sufficient for achieving vitrification with relatively low concentrations of CPA.

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...

Design of Refractory Metal Heat Pipe Life Test Environment Chamber, Cooling System, and Radio Frequency Heating System

NASA Technical Reports Server (NTRS)

Martin, J. J.; Bragg-Sitton, S. M.; Reid, R. S.; Stewart, E. T.; Davis, J. D.

2011-01-01

A series of 16 Mo-44.5%Re alloy/sodium heat pipes will be experimentally tested to examine heat pipe aging. To support this evaluation, an environmental test chamber and a number of auxiliary subsystems are required. These subsystems include radio frequency (RF) power supplies/inductive coils, recirculation water coolant loops, and chamber gas conditioning. The heat pipes will be grouped, based on like power and gas mixture requirements, into three clusters of five units each, configured in a pentagonal arrangement. The highest powered heat pipe will be tested separately. Test chamber atmospheric purity is targeted at <0.3 ppb oxygen at an approximate operating pressure of 76 torr (.1.5 psia), maintained by active purification (oxygen level is comparable to a 10(exp -6) torr environment). Treated water will be used in two independent cooling circuits to remove .85 kW. One circuit will service the RF hardware while the other will maintain the heat pipe calorimetry. Initial procedures for the startup and operation of support systems have been identified. Each of these subsystems is outfitted with a variety of instrumentation, integrated with distributed real-time controllers and computers. A local area network provides communication between all devices. This data and control network continuously monitors the health of the test hardware, providing warning indicators followed by automatic shutdown should potentially damaging conditions develop. During hardware construction, a number of checkout tests.many making use of stainless steel prototype heat pipes that are already fabricated.will be required to verify operation.

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.

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...

Titanium based flat heat pipes for computer chip cooling

NASA Astrophysics Data System (ADS)

Soni, Gaurav; Ding, Changsong; Sigurdson, Marin; Bozorgi, Payam; Piorek, Brian; MacDonald, Noel; Meinhart, Carl

2008-11-01

We are developing a highly conductive flat heat pipe (called Thermal Ground Plane or TGP) for cooling computer chips. Conventional heat pipes have circular cross sections and thus can't make good contact with chip surface. The flatness of our TGP will enable conformal contact with the chip surface and thus enhance cooling efficiency. Another limiting factor in conventional heat pipes is the capillary flow of the working fluid through a wick structure. In order to overcome this limitation we have created a highly porous wick structure on a flat titanium substrate by using micro fabrication technology. We first etch titanium to create very tall micro pillars with a diameter of 5 μm, a height of 40 μm and a pitch of 10 μm. We then grow a very fine nano structured titania (NST) hairs on all surfaces of the pillars by oxidation in H202. In this way we achieve a wick structure which utilizes multiple length scales to yield high performance wicking of water. It's capable of wicking water at an average velocity of 1 cm/s over a distance of several cm. A titanium cavity is laser-welded onto the wicking substrate and a small quantity of water is hermetically sealed inside the cavity to achieve a TGP. The thermal conductivity of our preliminary TGP was measured to be 350 W/m-K, but has the potential to be several orders of magnitude higher.

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.

Heat Pipe Systems

NASA Technical Reports Server (NTRS)

1993-01-01

The heat pipe was developed to alternately cool and heat without using energy or any moving parts. It enables non-rotating spacecraft to maintain a constant temperature when the surface exposed to the Sun is excessively hot and the non Sun-facing side is very cold. Several organizations, such as Tropic-Kool Engineering Corporation, joined NASA in a subsequent program to refine and commercialize the technology. Heat pipes have been installed in fast food restaurants in areas where humid conditions cause materials to deteriorate quickly. Moisture removal was increased by 30 percent in a Clearwater, FL Burger King after heat pipes were installed. Relative humidity and power consumption were also reduced significantly. Similar results were recorded by Taco Bell, which now specifies heat pipe systems in new restaurants in the Southeast.

SP-100 system thaw and start-up strategies

NASA Astrophysics Data System (ADS)

Kirpich, A.; Choe, H.

The authors review several strategies that have been considered during calendar year 1990 for SP-100 system start-up and thaw, and provide results of screening analyses. Screening studies have identified three concepts which are capable of thawing the SP-100 system: (1) a heat pipe approach in which reactor-heated heat pipes are routed to equipment enclosed within a thaw cavity; (2) a bleed-tube approach in which perforated tubes routed through ducting and components achieve thaw progressively as the result of successively thawing bleed-holes; and (3) a NaK traceline approach in which reactor-heated NaK is routed along ducting and components and achieves lithium thaw by both radiative and conductive coupling methods. Key issues have been identified in connection with pump start-up, reactor and power converter transient behavior, and hydraulic circuit stability. Pump start-up stability is accomplished by a linked-pump hydraulic arrangement.

Heat pipes for terrestrial applications in dehumidification systems

NASA Technical Reports Server (NTRS)

Khattar, Mukesh K.

1988-01-01

A novel application of heat pipes which greatly enhances dehumidification performance of air-conditioning systems is presented. When an air-to-air heat pipe heat exchanger is placed between the warm return air and cold supply air streams of an air conditioner, heat is efficiently transferred from the return air to the supply air. As the warm return air precools during this process, it moves closer to its dew-point temperature. Therefore, the cooling system works less to remove moisture. This paper discusses the concept, its benefits, the challenges of incorporating heat pipes in an air-conditioning system, and the preliminary results from a field demonstration of an industrial application.

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.

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.

Convection's enhancement in thermal micro pipes using extra fluid and shape memory material

NASA Astrophysics Data System (ADS)

Mihai, Ioan; Sprinceana, Siviu

2016-12-01

Up to now, there have been developed various applications of thermal micro pipes[1-3], such as refrigerating systems, high heat flux electronics cooling, and biological devices etc., based on vacuum vaporization followed by a convective phenomenon that allows vapor transfer from the vaporization area to the condensation one. This article presents studies carried out on the enhancement of the convective phenomenon taking place in flat thermal micro pipes. The proposed method[4] is aimed at the cooling of power electronics components, such as microprocessors. The conducted research focused on the use of shape memory materials that allow, by a semi-active method, to bring extra fluid in the vaporization area of the thermal micro pipe. The conducted investigations analyzed the variation of the liquid layer thickness in the trapezoidal micro channels and the thermal flow change over time. The modification of liquid flow was studied in correlation with the capacity of the polysynthetic material to retain the most extra fluid in its pores. The enhancement of the convective heat transfer phenomenon in flat thermal micro pipes was investigated in correspondence to the increase of liquid quantity in the vaporization zone. The charts obtained by aid of Mathcad[5] allowed to represent the evolution during a period of time (or with the pipe's length) of the liquid film thickness, the flow and the thermal flow, as a function of the liquid supply variation due to the shape memory materials and the modification of the working temperature.

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.

46 CFR 56.85-10 - Preheating.

Code of Federal Regulations, 2012 CFR

2012-10-01

.... (9) Heating rate for furnace, gas, electric resistance, and other surface heating methods must not... thickness in inches for thickness over 2 inches. (10) Heating route for induction heating must not exceed... still air. When furnace cooling is used, the pipe sections must be cooled in the furnace to 1000 °F and...

46 CFR 56.85-10 - Preheating.

Code of Federal Regulations, 2010 CFR

2010-10-01

.... (9) Heating rate for furnace, gas, electric resistance, and other surface heating methods must not... thickness in inches for thickness over 2 inches. (10) Heating route for induction heating must not exceed... still air. When furnace cooling is used, the pipe sections must be cooled in the furnace to 1000 °F and...

46 CFR 56.85-10 - Preheating.

Code of Federal Regulations, 2014 CFR

2014-10-01

.... (9) Heating rate for furnace, gas, electric resistance, and other surface heating methods must not... thickness in inches for thickness over 2 inches. (10) Heating route for induction heating must not exceed... still air. When furnace cooling is used, the pipe sections must be cooled in the furnace to 1000 °F and...

46 CFR 56.85-10 - Preheating.

Code of Federal Regulations, 2013 CFR

2013-10-01

.... (9) Heating rate for furnace, gas, electric resistance, and other surface heating methods must not... thickness in inches for thickness over 2 inches. (10) Heating route for induction heating must not exceed... still air. When furnace cooling is used, the pipe sections must be cooled in the furnace to 1000 °F and...

46 CFR 56.85-10 - Preheating.

Code of Federal Regulations, 2011 CFR

2011-10-01

.... (9) Heating rate for furnace, gas, electric resistance, and other surface heating methods must not... thickness in inches for thickness over 2 inches. (10) Heating route for induction heating must not exceed... still air. When furnace cooling is used, the pipe sections must be cooled in the furnace to 1000 °F and...

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.

Gravity flow rate of solids through orifices and pipes

NASA Technical Reports Server (NTRS)

Gardner, J. F.; Smith, J. E.; Hobday, J. M.

1977-01-01

Lock-hopper systems are the most common means for feeding solids to and from coal conversion reactor vessels. The rate at which crushed solids flow by gravity through the vertical pipes and valves in lock-hopper systems affects the size of pipes and valves needed to meet the solids-handling requirements of the coal conversion process. Methods used to predict flow rates are described and compared with experimental data. Preliminary indications are that solids-handling systems for coal conversion processes are over-designed by a factor of 2 or 3.

GPM Avionics Module Heat Pipes Design and Performance Test Results

NASA Technical Reports Server (NTRS)

Ottenstein, Laura; DeChristopher, Mike

2011-01-01

The Global Precipitation Measurement (GPM) mission is an international network of satellites that provide the next-generation global observations of rain and snow. The GPM core satellite carries an advanced radar / radiometer system to measure precipitation from space and serve as a reference standard to unify precipitation measurements from a constellation of research and operational satellites. Through improved measurements of precipitation globally, the GPM mission will help to advance our understanding of Earth's water and energy cycle, improve forecasting of extreme events that cause natural hazards and disasters, and extend current capabilities in using accurate and timely information of precipitation to directly benefit society. The avionics module on the core satellite contains a number of electronics boxes which are cooled by a network of aluminum/ammonia heat pipes and a honeycomb radiator which contains thirteen embedded aluminum/ammonia heat pipes. All heat pipes were individually tested by the vendor (Advanced Cooling Technologies, Inc.) prior to delivery. Following delivery to NASA, the flight avionics radiator and the flight spare transport heat pipes were mounted to flight-like test structure and a system level thermal vacuum test was performed. This test, which used simulators in place of all electronics boxes, was done to verify the operation of the thermal control system as a whole. This presentation will discuss the design of the avionics module heat pipes, and then discuss performance tests results for the individual heat pipes prior to delivery and for the system level thermal vacuum test. All heat pipes met their performance requirements. However, it was found that the power was too low in some instances to start all of the smaller radiator spreader heat pipes when they were tested in a reflux configuration (which is the nominal test configuration). Although this lowered the efficiency of the radiator somewhat, it did not impact the operating temperatures of the electronics boxes.

Final summary report for 1989 inservice inspection (ISI) of SRS (Savannah River Site) 100-P Reactor tank

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

Morrison, J.M.; Loibl, M.W.

1989-12-15

The integrity of the SRS reactor tanks is a key factor affecting their suitability for continued service since, unlike the external piping system and components, the tanks are virtually irreplaceable. Cracking in various areas of the process water piping systems has occurred beginning in 1960 as a result of several degradation mechanisms, chiefly intergranular stress corrosion cracking (IGSCC) and chloride-induced transgranular cracking. IGSCC, currently the primary degradation mechanism, also occurred in the knuckle'' region (tank wall-to-bottom tube sheet transition piece) unique to C Reactor and was eventually responsible for that reactor being deactivated in 1985. A program of visual examinationsmore » of the SRS reactor tanks was initiated in 1968, which used a specially designed immersible periscope. Under that program the condition of the accessible tank welds and associated heat affected zones (HAZ) was evaluated on a five-year frequency. Prior to 1986, the scope of these inspections comprised approximately 20 percent of the accessible weld area. In late 1986 and early 1987 the scope of the inspections was expanded and a 100 percent visual inspection of accessible welds was performed of the P-, L-, and K-Reactor tanks. Supplemental dye penetrant examinations were performed in L Reactor on selected areas which showed visual indications. No evidence of cracking was detected in any of these inspections of the P-, L-, and K-Reactor tanks. 17 refs., 7 figs.« less

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.

Floating Loop System For Cooling Integrated Motors And Inverters Using Hot Liquid Refrigerant

DOEpatents

Hsu, John S [Oak Ridge, TN; Ayers, Curtis W [Kingston, TN; Coomer, Chester [Knoxville, TN; Marlino, Laura D [Oak Ridge, TN

2006-02-07

A floating loop vehicle component cooling and air-conditioning system having at least one compressor for compressing cool vapor refrigerant into hot vapor refrigerant; at least one condenser for condensing the hot vapor refrigerant into hot liquid refrigerant by exchanging heat with outdoor air; at least one floating loop component cooling device for evaporating the hot liquid refrigerant into hot vapor refrigerant; at least one expansion device for expanding the hot liquid refrigerant into cool liquid refrigerant; at least one air conditioning evaporator for evaporating the cool liquid refrigerant into cool vapor refrigerant by exchanging heat with indoor air; and piping for interconnecting components of the cooling and air conditioning system.

PBF Reactor Building (PER620). Camera faces southeast. Concrete placement will ...

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

PBF Reactor Building (PER-620). Camera faces southeast. Concrete placement will leave opening for neutron camera to be installed later. Note vertical piping within rebar. Photographer: John Capek. Date: July 6, 1967. INEEL negative no. 67-3514 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

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

Heat Rejection from a Variable Conductance Heat Pipe Radiator Panel

NASA Technical Reports Server (NTRS)

Jaworske, D. A.; Gibson, M. A.; Hervol, D. S.

2012-01-01

A titanium-water heat pipe radiator having an innovative proprietary evaporator configuration was evaluated in a large vacuum chamber equipped with liquid nitrogen cooled cold walls. The radiator was manufactured by Advanced Cooling Technologies, Inc. (ACT), Lancaster, PA, and delivered as part of a Small Business Innovative Research effort. The radiator panel consisted of five titanium-water heat pipes operating as thermosyphons, sandwiched between two polymer matrix composite face sheets. The five variable conductance heat pipes were purposely charged with a small amount of non-condensable gas to control heat flow through the condenser. Heat rejection was evaluated over a wide range of inlet water temperature and flow conditions, and heat rejection was calculated in real-time utilizing a data acquisition system programmed with the Stefan-Boltzmann equation. Thermography through an infra-red transparent window identified heat flow across the panel. Under nominal operation, a maximum heat rejection value of over 2200 Watts was identified. The thermal vacuum evaluation of heat rejection provided critical information on understanding the radiator s performance, and in steady state and transient scenarios provided useful information for validating current thermal models in support of the Fission Power Systems Project.

Pressure suppression containment system for boiling water reactor

DOEpatents

Gluntz, Douglas M.; Nesbitt, Loyd B.

1997-01-01

A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs.

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

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.

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.

Wind tunnel data of the analysis of heat pipe and wind catcher technology for the built environment

PubMed Central

Calautit, John Kaiser; Chaudhry, Hassam Nasarullah; Hughes, Ben Richard

2015-01-01

The data presented in this article were the basis for the study reported in the research articles entitled ‘Climate responsive behaviour heat pipe technology for enhanced passive airside cooling’ by Chaudhry and Hughes [10] which presents the passive airside cooling capability of heat pipes in response to gradually varying external temperatures and related to the research article “CFD and wind tunnel study of the performance of a uni-directional wind catcher with heat transfer devices” by Calautit and Hughes [1] which compares the ventilation performance of a standard roof mounted wind catcher and wind catcher incorporating the heat pipe technology. Here, we detail the wind tunnel test set-up and inflow conditions and the methodologies for the transient heat pipe experiment and analysis of the integration of heat pipes within the control domain of a wind catcher design. PMID:26958604

A thin gold coated hydrogen heat pipe-cryogenic target for external experiments at COSY

NASA Astrophysics Data System (ADS)

Abdel-Bary, M.; Abdel-Samad, S.; Elawadi, G. A.; Kilian, K.; Ritman, J.

2009-05-01

A gravity assisted Gold coated heat pipe (GCHP) with 5-mm diameter has been developed and tested to cool a liquid hydrogen target for external beam experiments at COSY. The need for a narrow target diameter leads us to study the effect of reducing the heat pipe diameter to 5 mm instead of 7 mm, to study the effect of coating the external surface of the heat pipe by a shiny gold layer (to decrease the radiation heat load), and to study the effect of using the heat pipe without using 20 layers of' super-insulation around it (aluminized Mylar foil) to keep the target diameter as small as possible. The developed gold coated heat pipe was tested with 20 layers of super-insulation (WI) and without super-insulation (WOI). The operating characteristics for both conditions were compared to show the advantages and disadvantages.

Water NSTF Design, Instrumentation, and Test Planning

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

Lisowski, Darius D.; Gerardi, Craig D.; Hu, Rui

The following report serves as a formal introduction to the water-based Natural convection Shutdown heat removal Test Facility (NSTF) program at Argonne. Since 2005, this US Department of Energy (DOE) sponsored program has conducted large scale experimental testing to generate high-quality and traceable validation data for guiding design decisions of the Reactor Cavity Cooling System (RCCS) concept for advanced reactor designs. The most recent facility iteration, and focus of this report, is the operation of a 1/2 scale model of a water-RCCS concept. Several features of the NSTF prototype align with the conceptual design that has been publicly released formore » the AREVA 625 MWt SC-HTGR. The design of the NSTF also retains all aspects common to a fundamental boiling water thermosiphon, and thus is well poised to provide necessary experimental data to advance basic understanding of natural circulation phenomena and contribute to computer code validation. Overall, the NSTF program operates to support the DOE vision of aiding US vendors in design choices of future reactor concepts, advancing the maturity of codes for licensing, and ultimately developing safe and reliable reactor technologies. In this report, the top-level program objectives, testing requirements, and unique considerations for the water cooled test assembly are discussed, and presented in sufficient depth to support defining the program’s overall scope and purpose. A discussion of the proposed 6-year testing program is then introduced, which outlines the specific strategy and testing plan for facility operations. The proposed testing plan has been developed to meet the toplevel objective of conducting high-quality test operations that span across a broad range of single- and two-phase operating conditions. Details of characterization, baseline test cases, accident scenario, and parametric variations are provided, including discussions of later-stage test cases that examine the influence of geometric variations and off-normal configurations. The facility design follows, including as-built dimensions and specifications of the various mechanical and liquid systems, design choices for the test section, water storage tank, and network piping. Specifications of the instrumentation suite are then presented, along with specific information on performance windows, measurement uncertainties, and installation locations. Finally, descriptions of the control systems and heat removal networks are provided, which have been engineered to support precise quantification of energy balances and facilitate well-controlled test operations.« less

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.

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.

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

Cooling system for continuous metal casting machines

DOEpatents

Draper, Robert; Sumpman, Wayne C.; Baker, Robert J.; Williams, Robert S.

1988-01-01

A continuous metal caster cooling system is provided in which water is supplied in jets from a large number of small nozzles 19 against the inner surface of rim 13 at a temperature and with sufficient pressure that the velocity of the jets is sufficiently high that the mode of heat transfer is substantially by forced convection, the liquid being returned from the cooling chambers 30 through return pipes 25 distributed interstitially among the nozzles.

Cooling system for continuous metal casting machines

DOEpatents

Draper, R.; Sumpman, W.C.; Baker, R.J.; Williams, R.S.

1988-06-07

A continuous metal caster cooling system is provided in which water is supplied in jets from a large number of small nozzles against the inner surface of rim at a temperature and with sufficient pressure that the velocity of the jets is sufficiently high that the mode of heat transfer is substantially by forced convection, the liquid being returned from the cooling chambers through return pipes distributed interstitially among the nozzles. 9 figs.

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.

Welding High Strength Modern Line Pipe Steel

NASA Astrophysics Data System (ADS)

Goodall, Graeme Robertson

The effect of modern mechanized girth welding on high strength line pipe has been investigated. The single cycle grain coarsened heat affected zone in three grade 690 line pipe steels and a grade 550 steel has been simulated using a Gleeble thermo-mechanical simulator. The continuous cooling transformation diagrams applicable to the grain coarsened heat affected zone resulting from a range of heat inputs applicable to modern mechanized welding have been established by dilatometry and metallography. The coarse grained heat affected zone was found to transform to lath martensite, bainite, and granular bainite depending on the cooling rate. The impact toughness of the steels was measured using Charpy impact toughness and compared to the toughness of the grain coarsened heat affected zone corresponding to a welding thermal cycle. The ductile to brittle transition temperature was found to be lowest for the steel with the highest hardenability. The toughness resulting from three different thermal cycles including a novel interrupted intercritically reheated grain coarsened (NTR ICR GC HAZ) that can result from dual torch welding at fast travel speed and close torch spacing have been investigated. All of the thermally HAZ regions showed reduced toughness that was attributed to bainitic microstructure and large effective grain sizes. Continuous cooling transformation diagrams for five weld metal chemistries applicable to mechanized pulsed gas metal arc welding of modern high strength pipe steel (SMYS>550 MPa) have been constructed. Welds at heat inputs of 1.5 kJmm-1 and 0.5 kJmm-1 have been created for simulation and analysis. Dilatometric analysis was performed on weld metal specimens cut from single pass 1.5 kJmm-1 as deposited beads. The resulting microstructures were found to range from martensite to polygonal ferrite. There is excellent agreement between the simulated and as deposited weld metal regions. Toughness testing indicates improved energy absorption at -20 °C with increased cooling time.

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

Chattopadhyay, J.; Dutta, B.K.; Kushwaha, H.S.

Leak-Before-Break (LBB) is being used to design the primary heat transport piping system of 500 MWe Indian Pressurized Heavy Water Reactors (IPHWR). The work is categorized in three directions to demonstrate three levels of safety against sudden catastrophic break. Level 1 is inherent in the design procedure of piping system as per ASME Sec.III with a well defined factor of safety. Level 2 consists of fatigue crack growth study of a postulated part-through flaw at the inside surface of pipes. Level 3 is stability analysis of a postulated leakage size flaw under the maximum credible loading condition. Developmental work relatedmore » to demonstration of level 2 and level 3 confidence is described in this paper. In a case study on fatigue crack growth on PHT straight pipes for level 2, negligible crack growth is predicted for the life of the reactor. For level 3 analysis, the R6 method has been adopted. A database to evaluate SIF of elbows with throughwall flaws under combined internal pressure and bending moment has been generated to provide one of the inputs for R6 method. The methodology of safety assessment of elbow using R6 method has been demonstrated for a typical pump discharge elbow. In this analysis, limit load of the cracked elbow has been determined by carrying out elasto-plastic finite element analysis. The limit load results compared well with those given by Miller. However, it requires further study to give a general form of limit load solution. On the experimental front, a set of small diameter pipe fracture experiments have been carried out at room temperature and 300{degrees}C. Two important observations of the experiments are - appreciable drop in maximum load at 300{degrees}C in case of SS pipes and out-of-plane crack growth in case of CS pipes. Experimental load deflection curves are finally compared with five J-estimation schemes predictions. A material database of PHT piping materials is also being generated for use in LBB analysis.« less

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

Impervious surfaces and sewer pipe effects on stormwater runoff temperature

NASA Astrophysics Data System (ADS)

Sabouri, F.; Gharabaghi, B.; Mahboubi, A. A.; McBean, E. A.

2013-10-01

The warming effect of the impervious surfaces in urban catchment areas and the cooling effect of underground storm sewer pipes on stormwater runoff temperature are assessed. Four urban residential catchment areas in the Cities of Guelph and Kitchener, Ontario, Canada were evaluated using a combination of runoff monitoring and modelling. The stormwater level and water temperature were monitored at 10 min interval at the inlet of the stormwater management ponds for three summers 2009, 2010 and 2011. The warming effect of the ponds is also studied, however discussed in detail in a separate paper. An artificial neural network (ANN) model for stormwater temperature was trained and validated using monitoring data. Stormwater runoff temperature was most sensitive to event mean temperature of the rainfall (EMTR) with a normalized sensitivity coefficient (Sn) of 1.257. Subsequent levels of sensitivity corresponded to the longest sewer pipe length (LPL), maximum rainfall intensity (MI), percent impervious cover (IMP), rainfall depth (R), initial asphalt temperature (AspT), pipe network density (PND), and rainfall duration (D), respectively. Percent impervious cover of the catchment area (IMP) was the key parameter that represented the warming effect of the paved surfaces; sensitivity analysis showed IMP increase from 20% to 50% resulted in runoff temperature increase by 3 °C. The longest storm sewer pipe length (LPL) and the storm sewer pipe network density (PND) are the two key parameters that control the cooling effect of the underground sewer system; sensitivity analysis showed LPL increase from 345 to 966 m, resulted in runoff temperature drop by 2.5 °C.

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.

Twenty years of experience in monitoring 41Ar in a research reactor and decrease of its discharge into the environment.

PubMed

Fukui, M

2004-04-01

The radioactive gas 41Ar has been produced at high concentration by neutron activation near the reactor core in the Kyoto University Research Reactor. A pipe line for an exhaust stream, so-called sweep gas, was fabricated at the construction of the reactor in 1964 in order to exhale 41Ar from the facilities above to the environment. Other exhaust lines with decay tanks were established separately from the sweep line for both the cold neutron source in 1986 and the heavy-water tank in 1996, respectively, because a higher amount of 41Ar was thought to be produced from these facilities due to the improvement. As a result, a slight change in the flow rate of the exhaust was found to have a great deal of influence on both the 41Ar concentration in the reactor room and the rate of emission from the stack. By monitoring the exhaust air from the decay tanks, the mechanism for decreasing the emission was clarified together with identifying an obstacle, i.e., the condensate against the steady state flow, formed in the exhaust pipe. By setting the flow rate suitably in the exhaust line, the rate of 41Ar emission from the biological shielding into both the work place in the reactor room and the environment has been controlled as low as reasonably achievable.

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

NASA Technical Reports Server (NTRS)

El-Genk, Mohamed S. (Editor); Hoover, Mark D. (Editor)

1991-01-01

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects.

Space nuclear system expansion joints

NASA Technical Reports Server (NTRS)

Whitaker, W. D.; Shimazki, T. T.

1973-01-01

The engineering, design, and fabrication status of the expansion joint unit (EJU) to be employed in the NaK primary coolant piping loop of the 5-kwe Reactor thermoelectric system are described. Four EJU's are needed in the NaK primary coolant piping loop. The four EJU's which will be identical, utilize bellows as the flexing member, are hermetically sealed, and provide double containment. The bellows are of a nested-formed design, and are to be constructed of 1-ply thickness of 0.010-in. Inconel 718. The EJU's provide a minimum piping load margin of safety of +0.22.

Long Duration Exposure Facility (LDEF) low-temperature Heat Pipe Experiment Package (HEPP) flight results

NASA Technical Reports Server (NTRS)

Mcintosh, Roy; Mccreight, Craig; Brennan, Patrick J.

1992-01-01

The Low Temperature Heat Pipe Flight Experiment (HEPP) is a fairly complicated thermal control experiment that was designed to evaluate the performance of two different low temperature ethane heat pipes and a n-Heptane Phase Change Material (PCM) canister. A total of 388 days of continuous operation with an axially grooved aluminum fixed conductance heat pipe of axially grooved stainless steel heat pipe diode was demonstrated before the EDS batteries lost power. The inability of the HEPP's radiator to cool below 190 K in flight prevented freezing of the PCM and the opportunity to conduct transport tests with the heat pipes. Post flight tests showed that the heat pipes and the PCM are still functioning. This paper presents a summary of the flight data analysis for the HEPP and its related support systems. Pre and post-flight thermal vacuum tests results are presented for the HEPP thermal control system along with individual heat pipe performance and PCM behavior. Appropriate SIG related systems data will also be included along with a 'lessons learned' summary.

Fuel leak detection apparatus for gas cooled nuclear reactors

DOEpatents

Burnette, Richard D.

1977-01-01

Apparatus is disclosed for detecting nuclear fuel leaks within nuclear power system reactors, such as high temperature gas cooled reactors. The apparatus includes a probe assembly that is inserted into the high temperature reactor coolant gaseous stream. The probe has an aperture adapted to communicate gaseous fluid between its inside and outside surfaces and also contains an inner tube for sampling gaseous fluid present near the aperture. A high pressure supply of noncontaminated gas is provided to selectively balance the pressure of the stream being sampled to prevent gas from entering the probe through the aperture. The apparatus includes valves that are operable to cause various directional flows and pressures, which valves are located outside of the reactor walls to permit maintenance work and the like to be performed without shutting down the reactor.

System Study: Reactor Core Isolation Cooling 1998-2014

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

Schroeder, John Alton

2015-12-01

This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

Experimental investigation of a new method for advanced fast reactor shutdown cooling

NASA Astrophysics Data System (ADS)

Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

2017-07-01

We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

Improving the Parametric Method of Cost Estimating Relationships of Naval Ships

DTIC Science & Technology

2014-06-01

tool since the total cost of the ship is broken down into smaller parts as defined by the WBS. The Navy currently uses the Expanded Ship Work Breakdown...Includes boilers , reactors, turbines, gears, shafting, propellers, steam piping, lube oil piping, and radiation 300 Electric Plant Includes ship...spaces, ladders, storerooms, laundry, and workshops 700 Armament Includes guns, missile launchers, ammunition handling and stowage, torpedo tubes , depth

Transient Load Following and Control Analysis of Advanced S-CO2 Power Conversion with Dry Air Cooling

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

Moisseytsev, Anton; Sienicki, James J.

2016-01-01

Supercritical carbon dioxide (S-CO2) Brayton cycles are under development as advanced energy converters for advanced nuclear reactors, especially the Sodium-Cooled Fast Reactor (SFR). The use of dry air cooling for direct heat rejection to the atmosphere ultimate heat sink is increasingly becoming a requirement in many regions due to restrictions on water use. The transient load following and control behavior of an SFR with an S-CO2 cycle power converter utilizing dry air cooling have been investigated. With extension and adjustment of the previously existing control strategy for direct water cooling, S-CO2 cycle power converters can also be used for loadmore » following operation in regions where dry air cooling is a requirement« less

Heat pipe thermal conditioning panel

NASA Technical Reports Server (NTRS)

Saaski, E. W.; Loose, J. D.; Mccoy, K. E.

1974-01-01

Thermal control of electronic hardware and experiments on future space vehicles is critical to proper functioning and long life. Thermal conditioning panels (cold plates) are a baseline control technique in current conceptual studies. Heat generating components mounted on the panels are typically cooled by fluid flowing through integral channels within the panel. However, replacing the pumped fluid coolant loop within the panel with heat pipes offers attractive advantages in weight, reliability, and installation. This report describes the development and fabrication of two large 0.76 x 0.76 m heat pipe thermal conditioning panels to verify performance and establish the design concept.

Developments and Tendencies in Fission Reactor Concepts

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Fuji-Ie, Y.

This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC) - as an advanced and promising reactor system that offers solutions to the above problems. The difference (not confrontation) between the approaches to nuclear power development based on the principles of “inherent safety” and “natural safety” is demonstrated.

WATER PROCESS SYSTEM FLOW DIAGRAM FOR MTR, TRA603. SUMMARY OF ...

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

WATER PROCESS SYSTEM FLOW DIAGRAM FOR MTR, TRA-603. SUMMARY OF COOLANT FLOW FROM WORKING RESERVOIR TO INTERIOR OF REACTOR'S THERMAL SHIELD. NAMES TANK SECTIONS. PIPE AND DRAIN-LINE SIZES. SHOWS DIRECTION OF AIR FLOW THROUGH PEBBLE AND GRAPHITE BLOCK ZONE. NEUTRON CURTAIN AND THERMAL COLUMN DOOR. BLAW-KNOX 3150-92-7, 3/1950. INL INDEX NO. 531-0603-51-098-100036, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Pressurizer tank upper support

DOEpatents

Baker, Tod H.; Ott, Howard L.

1994-01-01

A pressurizer tank in a pressurized water nuclear reactor is mounted between structural walls of the reactor on a substructure of the reactor, the tank extending upwardly from the substructure. For bearing lateral loads such as seismic shocks, a girder substantially encircles the pressurizer tank at a space above the substructure and is coupled to the structural walls via opposed sway struts. Each sway strut is attached at one end to the girder and at an opposite end to one of the structural walls, and the sway struts are oriented substantially horizontally in pairs aligned substantially along tangents to the wall of the circular tank. Preferably, eight sway struts attach to the girder at 90.degree. intervals. A compartment encloses the pressurizer tank and forms the structural wall. The sway struts attach to corners of the compartment for maximum stiffness and load bearing capacity. A valve support frame carrying the relief/discharge piping and valves of an automatic depressurization arrangement is fixed to the girder, whereby lateral loads on the relief/discharge piping are coupled directly to the compartment rather than through any portion of the pressurizer tank. Thermal insulation for the valve support frame prevents thermal loading of the piping and valves. The girder is shimmed to define a gap for reducing thermal transfer, and the girder is free to move vertically relative to the compartment walls, for accommodating dimensional variation of the pressurizer tank with changes in temperature and pressure.

Automatic Target Recognition Using Nonlinear Autoregressive Neural Networks

DTIC Science & Technology

2014-03-27

Lee and Chang (2009) employed a NARXNet for studying the thermodynamics in a pulsating heat pipe (PHP), a type of cooling device which contains...in Thermal Dynamics Identifcation of a Pulsating Heat Pipe . Energy Consversion and Management, 1069-1078. Lisboa, P. J. (2002). A review of...function by adjusting the values of the connections between elements. This flexibility allows ANNs to perform complex functions in fields to include

Quasi-2D Unsteady Flow Procedure for Real Fluids (PREPRINT)

DTIC Science & Technology

2006-05-17

water /steam/ oil piping networks, refinery systems, gas-turbine secondary flow -path and cooling networks...friction factor, f, which is a function of the local Reynolds number and the wall surface roughness . For the viscous flow examples presented below, the...3.5 4 4.5 Time ( s ) V el oc ity (m / s ) Line 2 Inlet 25% 50% 75% Exit Velocity Figure 4. Water transient viscous pipe flow using

Enhanced heat transfer with full circumferential ribs in helical pipe

NASA Astrophysics Data System (ADS)

Chang, S. W.; Su, L. M.; Yang, T. L.

2002-08-01

This paper describes an experimental study of heat transfers in the smooth-walled and rib-roughened helical pipes with reference to the design of enhanced cooling passages in the cylinder head and liner of a marine propulsive diesel engine. The manner in which the repeated ribs modify the forced heat convection in the helical pipe is considered for the case where the flow is turbulent upon entering the coil but laminar in further downstream. A selection of experimental results illustrates the individual and interactive effects of Dean vortices and rib-flows on heat transfer along the inner and outer helixes of coils. The experimental-based observations reveal that the centrifugal force modifies the heat transfer in a manner to generate circumferential heat transfer variation with better cooling performance on the outer edge relative to its inner counterpart even with the agitated flow field caused by the repeated ribs. Heat transfer augmentation factor in the range of 1.3 - 3 times of the smooth-walled level is achieved using the present ribbing geometry. A set of empirical correlations based on the experimental data has been developed to permit the evaluation of heat transfers along the inner and outer helixes of the smooth-walled and rib-roughened helical pipes.

Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

DOEpatents

Peterson, Per F.

2013-05-14

A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

Preparation of high temperature gas-cooled reactor fuel element

DOEpatents

Bradley, Ronnie A.; Sease, John D.

1976-01-01

This invention relates to a method for the preparation of high temperature gas-cooled reactor (HTGR) fuel elements wherein uncarbonized fuel rods are inserted in appropriate channels of an HTGR fuel element block and the entire block is inserted in an autoclave for in situ carbonization under high pressure. The method is particularly applicable to remote handling techniques.

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

Nuclear reactor insulation and preheat system

DOEpatents

Wampole, Nevin C.

1978-01-01

An insulation and preheat system for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the compartment. An external surface of the compartment or enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair.

A Comparison of Fission Power System Options for Lunar and Mars Surface Applications

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2006-01-01

This paper presents a comparison of reactor and power conversion design options for 50 kWe class lunar and Mars surface power applications with scaling from 25 to 200 kWe. Design concepts and integration approaches are provided for three reactor-converter combinations: gas-cooled Brayton, liquid-metal Stirling, and liquid-metal thermoelectric. The study examines the mass and performance of low temperature, stainless steel based reactors and higher temperature refractory reactors. The preferred system implementation approach uses crew-assisted assembly and in-situ radiation shielding via installation of the reactor in an excavated hole. As an alternative, self-deployable system concepts that use earth-delivered, on-board radiation shielding are evaluated. The analyses indicate that among the 50 kWe stainless steel reactor options, the liquid-metal Stirling system provides the lowest mass at about 5300 kg followed by the gas-cooled Brayton at 5700 kg and the liquid-metal thermoelectric at 8400 kg. The use of a higher temperature, refractory reactor favors the gas-cooled Brayton option with a system mass of about 4200 kg as compared to the Stirling and thermoelectric options at 4700 and 5600 kg, respectively. The self-deployed concepts with on-board shielding result in a factor of two system mass increase as compared to the in-situ shielded concepts.

DE-NE0008277_PROTEUS final technical report 2018

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

Enqvist, Andreas

This project details re-evaluations of experiments of gas-cooled fast reactor (GCFR) core designs performed in the 1970s at the PROTEUS reactor and create a series of International Reactor Physics Experiment Evaluation Project (IRPhEP) benchmarks. Currently there are no gas-cooled fast reactor (GCFR) experiments available in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). These experiments are excellent candidates for reanalysis and development of multiple benchmarks because these experiments provide high-quality integral nuclear data relevant to the validation and refinement of thorium, neptunium, uranium, plutonium, iron, and graphite cross sections. It would be cost prohibitive to reproduce suchmore » a comprehensive suite of experimental data to support any future GCFR endeavors.« less

Reactor core isolation cooling system

DOEpatents

Cooke, F.E.

1992-12-08

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

Reactor core isolation cooling system

DOEpatents

Cooke, Franklin E.

1992-01-01

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

COOLED NEUTRONIC REACTOR

DOEpatents

Binner, C.R.; Wilkie, C.B.

1958-03-18

This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

ETR HEAT EXCHANGER BUILDING, TRA644. EAST SIDE. CAMERA FACING WEST. ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. EAST SIDE. CAMERA FACING WEST. NOTE COURSE OF PIPE FROM GROUND AND FOLLOWING ROOF OF BUILDING. MTR BUILDING IN BACKGROUND AT RIGHT EDGE OF VIEW. INL NEGATIVE NO. HD46-36-3. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Low cost solar array project: Experimental process system development unit for producing semiconductor-grade silicon using the silane-to-silicon process

NASA Technical Reports Server (NTRS)

1981-01-01

The results of the free space reactor experimental work are summarized. Overall, the objectives were achieved and the unit can be confidently scaled to the EPSDU size based on the experimental work and supporting theoretical analyses. The piping and instrumentation of the fluidized bed reactor was completed.

PBF Reactor Building (PER620). Cubicle 10. Camera facing southeast. Loop ...

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

PBF Reactor Building (PER-620). Cubicle 10. Camera facing southeast. Loop pressurizer on right. Other equipment includes loop strained, control valves, loop piping, pressurizer interchanger, and cleanup system cooler. High-density shielding brick walls. Photographer: Kirsh. Date: November 2, 1970. INEEL negative no. 70-4908 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Pressure suppression containment system for boiling water reactor

DOEpatents

Gluntz, D.M.; Nesbitt, L.B.

1997-01-21

A system is disclosed for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs. 3 figs.

Conodont geothermometry in pyroclastic kimberlite: constraints on emplacement temperatures and cooling histories

NASA Astrophysics Data System (ADS)

Pell, Jennifer; Russell, James K.; Zhang, Shunxin

2018-03-01

Kimberlite pipes from Chidliak, Baffin Island, Nunavut, Canada host surface-derived Paleozoic carbonate xenoliths containing conodonts. Conodonts are phosphatic marine microfossils that experience progressive, cumulative and irreversible colour changes upon heating that are experimentally calibrated as a conodont colour alteration index (CAI). CAI values permit us to estimate the temperatures to which conodont-bearing rocks have been heated. Conodonts have been recovered from 118 samples from 89 carbonate xenoliths collected from 12 of the pipes and CAI values within individual carbonate xenoliths show four types of CAI distributions: (1) CAI values that are uniform throughout the xenolith; (2) lower CAIs in core of a xenolith than the rim; (3) CAIs that increase from one side of the xenolith to the other; and, (4) in one xenolith, higher CAIs in the xenolith core than at the rim. We have used thermal models for post-emplacement conductive cooling of kimberlite pipes and synchronous heating of conodont-bearing xenoliths to establish the temperature-time history of individual xenoliths within the kimberlite bodies. Model results suggest that the time-spans for xenoliths to reach the peak temperatures recorded by CAIs varies from hours for the smallest xenoliths to 2 or 3 years for the largest xenoliths. The thermal modelling shows the first three CAI patterns to be consistent with in situ conductive heating of the xenoliths coupled to the cooling host kimberlite. The fourth pattern remains an anomaly.

Potential Effects of Leak-Before-Break on Light Water Reactor Design.

DTIC Science & Technology

1985-08-26

Boiler and Pressure Vessel Code . In fact, section 3 of that code was created for nuclear applications. This... Boiler and Pressure Vessel Code . The only major change which leak-before-break would require in these analyses would be that all piping to be considered...XI of the ASME Boiler and Pressure Vessel Code , and is already required for all Class I piping systems in the plant. Class I systems are those

Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

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

Permana, Sidik; Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132; Sekimoto, Hiroshi

2010-12-23

Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period hasmore » been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.« less

Experimental simulation of latent heat thermal energy storage and heat pipe thermal transport for dish concentrator solar receiver

NASA Technical Reports Server (NTRS)

Narayanan, R.; Zimmerman, W. F.; Poon, P. T. Y.

1981-01-01

Test results on a modular simulation of the thermal transport and heat storage characteristics of a heat pipe solar receiver (HPSR) with thermal energy storage (TES) are presented. The HPSR features a 15-25 kWe Stirling engine power conversion system at the focal point of a parabolic dish concentrator operating at 827 C. The system collects and retrieves solar heat with sodium pipes and stores the heat in NaF-MgF2 latent heat storage material. The trials were run with a single full scale heat pipe, three full scale TES containers, and an air-cooled heat extraction coil to replace the Stirling engine heat exchanger. Charging and discharging, constant temperature operation, mixed mode operation, thermal inertial, etc. were studied. The heat pipe performance was verified, as were the thermal energy storage and discharge rates and isothermal discharges.

Startup analysis for a high temperature gas loaded heat pipe

NASA Technical Reports Server (NTRS)

Sockol, P. M.

1973-01-01

A model for the rapid startup of a high-temperature gas-loaded heat pipe is presented. A two-dimensional diffusion analysis is used to determine the rate of energy transport by the vapor between the hot and cold zones of the pipe. The vapor transport rate is then incorporated in a simple thermal model of the startup of a radiation-cooled heat pipe. Numerical results for an argon-lithium system show that radial diffusion to the cold wall can produce large vapor flow rates during a rapid startup. The results also show that startup is not initiated until the vapor pressure p sub v in the hot zone reaches a precise value proportional to the initial gas pressure p sub i. Through proper choice of p sub i, startup can be delayed until p sub v is large enough to support a heat-transfer rate sufficient to overcome a thermal load on the heat pipe.

Reactor safety method

DOEpatents

Vachon, Lawrence J.

1980-03-11

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

World Energy Data System (WENDS). Volume XI. Nuclear fission program summaries

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

Not Available

1979-06-01

Brief management and technical summaries of nuclear fission power programs are presented for nineteen countries. The programs include the following: fuel supply, resource recovery, enrichment, fuel fabrication, light water reactors, heavy water reactors, gas cooled reactors, breeder reactors, research and test reactors, spent fuel processing, waste management, and safety and environment. (JWR)

Breeder Reactors, Understanding the Atom Series.

ERIC Educational Resources Information Center

Mitchell, Walter, III; Turner, Stanley E.

The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

Cryogenic pellet production developments for long-pulse plasma operation

NASA Astrophysics Data System (ADS)

Meitner, S. J.; Baylor, L. R.; Combs, S. K.; Fehling, D. T.; McGill, J. M.; Duckworth, R. C.; McGinnis, W. D.; Rasmussen, D. A.

2014-01-01

Long pulse plasma operation on large magnetic fusion devices require multiple forms of cryogenically formed pellets for plasma fueling, on-demand edge localized mode (ELM) triggering, radiative cooling of the divertor, and impurity transport studies. The solid deuterium fueling and ELM triggering pellets can be formed by extrusions created by helium cooled, twin-screw extruder based injection system that freezes deuterium in the screw section. A solenoid actuated cutter mechanism is activated to cut the pellets from the extrusion, inserting them into the barrel, and then fired by the pneumatic valve pulse of high pressure gas. Fuel pellets are injected at a rate up to 10 Hz, and ELM triggering pellets are injected at rates up to 20 Hz. The radiative cooling and impurity transport study pellets are produced by introducing impurity gas into a helium cooled section of a pipe gun where it deposits in-situ. A pneumatic valve is opened and propellant gas is released downstream where it encounters a passive punch which initially accelerates the pellet before the gas flow around the finishes the pellet acceleration. This paper discusses the various cryogenic pellet production techniques based on the twin-screw extruder, pipe gun, and pellet punch designs.

Cryogenic pellet production developments for long-pulse plasma operation

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

Meitner, S. J.; Baylor, L. R.; Combs, S. K.

Long pulse plasma operation on large magnetic fusion devices require multiple forms of cryogenically formed pellets for plasma fueling, on-demand edge localized mode (ELM) triggering, radiative cooling of the divertor, and impurity transport studies. The solid deuterium fueling and ELM triggering pellets can be formed by extrusions created by helium cooled, twin-screw extruder based injection system that freezes deuterium in the screw section. A solenoid actuated cutter mechanism is activated to cut the pellets from the extrusion, inserting them into the barrel, and then fired by the pneumatic valve pulse of high pressure gas. Fuel pellets are injected at amore » rate up to 10 Hz, and ELM triggering pellets are injected at rates up to 20 Hz. The radiative cooling and impurity transport study pellets are produced by introducing impurity gas into a helium cooled section of a pipe gun where it deposits in-situ. A pneumatic valve is opened and propellant gas is released downstream where it encounters a passive punch which initially accelerates the pellet before the gas flow around the finishes the pellet acceleration. This paper discusses the various cryogenic pellet production techniques based on the twin-screw extruder, pipe gun, and pellet punch designs.« less

Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

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

Corradin, Michael; Anderson, M.; Muci, M.

This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintainmore » similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.« less

Influence of the ambient temperature on the cooling efficiency of the high performance cooling device with thermosiphon effect

NASA Astrophysics Data System (ADS)

Nemec, Patrik; Malcho, Milan

2018-06-01

This work deal with experimental measurement and calculation cooling efficiency of the cooling device working with a heat pipe technology. The referred device in the article is cooling device capable transfer high heat fluxes from electric elements to the surrounding. The work contain description, working principle and construction of cooling device. The main factor affected the dissipation of high heat flux from electronic elements through the cooling device to the surrounding is condenser construction, its capacity and option of heat removal. Experimental part describe the measuring method cooling efficiency of the cooling device depending on ambient temperature in range -20 to 40°C and at heat load of electronic components 750 W. Measured results are compared with results calculation based on physical phenomena of boiling, condensation and natural convection heat transfer.

Thermion: Verification of a thermionic heat pipe in microgravity

NASA Technical Reports Server (NTRS)

1991-01-01

The design and development is examined of a small excore heat pipe thermionic space nuclear reactor power system (SEHPTR). The need was identified for an in-space flight demonstration of a solar powered, thermionic heat pipe element. A demonstration would examine its performance and verify its operation in microgravity. The design of a microsatellite based technology demonstration experiment is proposed to measure the effects of microgravity on the performance of an integrated thermionic heat pipe device in low earth orbit. The specific objectives are to verify the operation of the liquid metal heat pipe and the cesium reservior in the space environment. Two design configurations are described; THERMION-I and THERMION-II. THERMION-I is designed for a long lifetime study of the operations of the thermionic heat pipe element in low earth orbit. Heat input to the element is furnished by a large mirror which collects solar energy and focuses it into a cavity containing the heat pipe device. THERMION-II is a much simpler device which is used for short term operation. This experiment remains attached to the Delta II second stage and uses energy from 500 lb of alkaline batteries to supply heat energy to the heat pipe device.

A new model for early Earth: heat-pipe cooling

NASA Astrophysics Data System (ADS)

Webb, A. G.; Moore, W. B.

2013-12-01

In the study of heat transport and lithospheric dynamics of early Earth, current models depend upon plate tectonic and vertical tectonic concepts. Plate tectonic models adequately account for regions with diverse lithologies juxtaposed along ancient shear zones, as seen at the famous Eoarchean Isua supracrustal belt of West Greenland. Vertical tectonic models to date have involved volcanism, sub- and intra-lithospheric diapirism, and sagduction, and can explain the geology of the best-preserved low-grade ancient terranes, such as the Paleoarchean Barberton and Pilbara greenstone belts. However, these models do not offer a globally-complete framework consistent with the geologic record. Plate tectonics models suggest that paired metamorphic belts and passive margins are among the most likely features to be preserved, but the early rock record shows no evidence of these terranes. Existing vertical tectonics models account for the >300 million years of semi-continuous volcanism and diapirism at Barberton and Pilbara, but when they explain the shearing record at Isua, they typically invoke some horizontal motion that cannot be differentiated from plate motion and is not a salient feature of the lengthy Barberton and Pilbara records. Despite the strengths of these models, substantial uncertainty remains about how early Earth evolved from magma ocean to plate tectonics. We have developed a new model, based on numerical simulations and analysis of the geologic record, that provides a coherent, global geodynamic framework for Earth's evolution from magma ocean to subduction tectonics. We hypothesize that heat-pipe cooling offers a viable mechanism for the lithospheric dynamics of early Earth. Our numerical simulations of heat-pipe cooling on early Earth indicate that a cold, thick, single-plate lithosphere developed as a result of frequent volcanic eruptions that advected surface materials downward. The constant resurfacing and downward advection caused compression as the surface rocks were forced radially inward, resulting in uplift, exhumation, and shortening. Declining heat sources over time led to an abrupt, dynamically spontaneous transition to plate tectonics. The model predicts a geological record with rapid, semi-continuous volcanic resurfacing; contractional deformation; a low geothermal gradient across the bulk of the lithosphere; and a rapid decrease in heat-pipe volcanism after the initiation of plate tectonics. Review of data from ancient cratons and the detrital zircon record is consistent with these predictions. In this presentation, we review these findings with a focus on comparison of the model predictions with the geologic record. This comparison suggests that Earth cooled via heat pipes until a ~3.2 Ga subduction initiation episode. The Isua record reflects long-lived contractional deformation, and the Barberton and Pilbara records preserve heat-pipe lithospheric development in regions without significant contraction. In summary, the heat-pipe model provides a view of early Earth that is more globally applicable than existing plate and vertical tectonic models.

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.

2005-01-01

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.; Hoover, Mark D.

1991-07-01

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects. (For individual items see A93-13752 to A93-13937)

PBF Reactor Building (PER620). Plot plan shows layout, including auxiliary ...

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

PBF Reactor Building (PER-620). Plot plan shows layout, including auxiliary buildings: Emergency Generator (621), Hose House (622), Cooling Tower Auxiliary (624), Maintenance and Storage Warehouse (625), Gas Cylinder Storage (627), Hose House (628), Cooling Tower (720), Substation (719), and other features. Road connections between PBF Reactor, its control building, and SPERT-I site. Note cable trenches along road to control building. Date: July 1965. Ebasco Services, PER-U-101. INEEL index no. 761-0100-00-205-123005 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID